Microstructure and hydrides in the near rolled-joint regions of ex-service CANDU pressure tube

Abstract The regions of CANDU® pressure tubes near the rolled joint (RJ) are the un-supported portions of the inlet and outlet ends just inboard of where the tubes make contact with the 403 stainless steel end fitting sections. Compared to the main body of the pressure tube, in-service conditions are more complex with large gradients in the flux, significant hydrogen isotope concentration profiles and varying operating temperatures. In order to increase the amount of data related to microstructure evolution, the microstructures near rolled-joint regions from an ex-service CANDU pressure tube were investigated. The results show that irradiation-induced effects on the microstructure are more significant than the temperature influence near the inlet rolled-joint region (that operates at lower temperatures). Thermal effects play a more dominant role than the flux near the outlet region (that operates at higher temperatures). Zirconium hydrides aligned in both circumferential and radial directions of the tube have been observed. The zirconium hydrides have an orientation relationship of (0002)αZr//( 1 ¯ 1 ¯ 1)ZrH, and [2 1 ¯ 1 ¯ 0]αZr//[011]ZrH with the α-Zr matrix; the hydride phases were identified as gamma (γ) hydride by observing superlattice reflections on the (110) planes in the electron diffraction patterns. The presence of γ hydrides at room temperature in the studied tube may have technical importance because in most of the previous research, δ hydrides were considered as the stable phase that would precipitate during cooling at reactor relevant rates. This assumption has been applied to both delayed hydride cracking (DHC) and fracture toughness experiments and during development of related models. It is critical to elucidate the phases of zirconium hydrides at relevant in-service conditions to enable representative experimental testing and modelling that can be used to verify and/or improve the predictive capability of material evolution and cracking behavior of zirconium alloys in nuclear reactors.