NSUF 24-5166: In-situ TEM study of irradiation effects in zirconium oxides

Zirconium alloys have been widely used as fuel cladding materials in light water nuclear reactors (LWR). In service, the inner cladding surfaces are progressively oxidized by oxygen containing species from the fuel pellets. The oxide layers formed on the inner cladding surface can degrade the thermal and mechanical properties of fuel rods and limit their lifetime, so it is important to understand their evolution under irradiation in order to improve the modelling of rod behavior in the reactor. In this project, we propose the use of in situ transmission electron microscopy (TEM) to conduct real-time investigations into the microstructural changes in zirconium oxides at various dose levels (0.5, 1, 10, and 20 dpa) and temperatures, specifically at 20°C, 360°C and 600°C (approximating the inner cladding surface temperature). This study aims to collect reliable data on radiation damage behavior in zirconium oxides, which includes the formation of defect clusters and the phase transformation between tetragonal/cubic-ZrO2 and monoclinic-ZrO2 at elevated temperatures and under different doses.

추가 정보

필드	값
Award Announced Date	2024-09-23T12:20:05.88
Awarded Institution	University of Wisconsin-Madison
Facility Tech Lead	Wei-Ying Chen
Irradiation Facility	Intermediate Voltage Electron Microscopy (IVEM)-Tandem Facility
PI	Junliang Liu
PI Email	[email protected]
Project Type	RTE
RTE Number	None