Damage Mechanisms - Failure analysis of cracked 304 stainless steel conduit of PWR incore instrumentation system

The incore instrumentation system of a nuclear reactor facilitates neutron flux mapping at specific core locations. Guide conduits extending from the seal table to the reactor pressure vessel guide and protect the movable incore neutron flux thimbles between the table and the lower reactor vessel head. Once filled with reactor coolant, the conduits become an extension to the reactor coolant system pressure boundary. The present work reports the examination of cracking detected in a TP304 stainless steel guide conduit in proximity to a weld at the upper surface of a TP304 seal table in a pressurized water reactor (PWR), resulting in the escape of reactor coolant, detected by the presence of boric acid deposits on the exterior of the conduit and table. Root cause analysis including dimensional measurements, chemical analysis, stereomicroscopy, metallography, and scanning electron microscopy showed that extensive cracking of the conduit and seal table material occurred due to stress corrosion cracking (SCC). Assessment showed that chlorine-containing deposits were present on the exterior of the conduit and on the surfaces of the seal table and were related to source(s) of chlorine in the seal table room in containment. Stainless steels are well understood to exhibit SCC susceptibility in environments with elevated temperatures, chloride contents, and increased tensile stress – particularly in non-post weld heat treated (PWHT) weld regions. However, chloride-induced SCC of such materials typically results in transgranular crack propagation, whereas the observed cracks were indicative of intergranular stress corrosion cracking (IGSCC). Microstructural analysis showed that the observed cracks occurred in sensitized areas of material in proximity to the weld, which resulted in increased susceptibility of the material to chloride-induced IGSCC.