Chemical interaction layer between uranium oxide fuel pellet and zirconium alloy cladding in pressurized water reactor

• Published in: Hé jìshū Vol. 46; no. 9; p. 090602
• Main Authors: WANG Huacai, YANG Dawei, CHENG Huanlin, TANG Qi, WANG Wei, QIAN Jin
• Format: Journal Article
• Language: Chinese
• Published: Science Press 01-09-2023
• Subjects: (u,zr)ox compound; chemical adhesion; chemical interaction layer; intact fuel rod; pressurized water reactor nuclear power plant
• DOI: 10.11889/j.0253-3219.2023.hjs.46.090602
• ISSN: 0253-3219

BackgroundDuring reactor operation, zirconium (Zr) alloy cladding is continuously oxidized as it gets in contact with fuel, and combines with the fuel to form a firm chemical interaction layer. This affects the thermal conductivity of the fuel gap, the mechanical properties of the cladding, and the mechanical interaction of the fuel cladding.PurposeThis work aims to obtain relevant analytical data on the chemical interaction layer between the irradiation Zr-alloy cladding and uranium oxide (UO2) pellets in a pressurized water reactor (PWR).MethodsFirst of all, the D13 intact fuel rod with a burnup of 45 GWd·tU-1 for PWRs in a nuclear power plant was chosen as UO2 pellets with a pellet enrichment of 4.45 wt%, and M5 Zr-alloy was used as the cladding materials. Then, a series of operations (cutting, pellet separation, inlaying, secondary cutting, inlaying, and polishing of cladding tubes) was conducted in the hot cell. The polished sample was transferred to the lead chamber and the UO2 fuel pellet was removed u