Oxidation products of Inconel alloys 600 and 690 in hydrogenated steam environments and their role in stress corrosion cracking

Inconel Alloys 600 and 690 are used extensively in components of Nuclear Pressurized Water Reactors (PWR) in the primary water loop which consists of H2 supersaturated steam. Alloy 600 has been found to crack intergranularly when exposed to primary water conditions. Alloy 690 was designed as a replacement and is generally regarded as immune to cracking. There is no consensus as to the mechanism which is responsible for cracking or the lack thereof in these alloys. In this work thermodynamic arguments for the stability of Ni and Cr compounds developed under pressurized water reactor environments ( PH2O and PH2 ) were experimentally tested. A mechanism is proposed to explain crack initiation and propagation alloy 600 along the grain boundaries where Cr2O3 has formed from the leaching of Cr from the matrix leaving behind a porous Ni-rich region. The mechanism is based on the thermodynamic potential for the transformation of a protective NiO surface layer into an amorphous non-protective Ni(OH)2 gel. This gel would also form along the grain boundaries and when hydrogenated steam reaches the porous Ni-rich regions. Crack initiation is then favored by tensile stressing of the grain boundary regions which can easily rupture the gelatinous film. The leaching of matrix Cr to form non-protective CrOOH gel at the crack tip followed by the exposure of fresh porous Ni to the environment also explains crack propagation in inconel alloy 600. The proposed crack initiation mechanism is not expected to occur in alloy 690 where a protective Cr2O 3 film covers the entire metal surface. However, crack propagation along the grain boundaries in alloy 600 and pre-cracked alloy 690 is expected to be active as hydroxide-forming reactions weaken the material at the grain boundaries.