[3A06-09] Fuel Cladding Materials 2

[3A06] The effect of applied tensile stress on microstructure evolution under high dose ion irradiation in Zry-2

*Luwei XUE¹, Hideo Watanabe¹ (1. KYUSHU UNIVERSITY)

Keywords:Zirconium alloy, Applied stress, Ion irradiation, Irradiation effects, Energy-dispersive spectroscopy

Zircaloy-2 is located closest to the fuel as the fuel cladding tube in BWRs. During reactor operation, fuel cladding tubes experience stress due to UO₂ pellet-cladding interactions. In this study, we investigated the microstructural evolution of Zircaloy-2 under 3.2 MeV Ni³⁺ ion irradiation using TEM examination and EDS analysis. The materials were irradiated at 300 °C and 400 °C under applied stress. At high-dose irradiation levels, loops only nucleated above a threshold dose of 20 dpa at 400 °C. The dissolution of Fe atoms from Zr(Fe, Cr)₂ precipitates due to irradiation was detected using EDS. However, the effect of stress on the dissolution rate of Fe-rich precipitates was minor.