Irradiation growth behaviour in zirconium-based alloy claddings for PWR

Kakiuchi, Kazuo ; Udagawa, Yutaka  ; Amaya, Masaki

In order to understand the dimensional stability of the fuel rod during long-term use in commercial LWRs, irradiation growth testing on various Zr-based fuel cladding materials including improved Zr-based alloys was conducted in the Halden reactor in Norway under prototypic irradiation and corrosion environment. Through detailed analysis using the experimental data, it was found that the addition of Nb in the improved alloy tends to reduce irradiation growth. In addition, the comparison of irradiation growth between the specimens with and without pre-charged hydrogen (about 200 and 400 ppm) showed that the effect of absorbed hydrogen on the acceleration of irradiation growth became significant when the hydrogen content exceeded the solubility limit of hydrogen at the corresponding irradiation temperature. Furthermore, based on the obtained data, the irradiation growth of the improved Zr-based alloys was formulated from an engineering viewpoint.