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U.S. Department of Energy
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Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods

Conference ·
OSTI ID:6980608

Twenty-one LWBR irradiation test rods containing ThO/sub 2/-UO/sub 2/ fuel and Zircaloy cladding with holes or cracks operated successfully. The irradiation tests were performed in the Engineering Test Reactor (ETR) and the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory and in the National Research Experimental Reactor (NRX) located at Chalk River. Zircaloy cladding corrosion on the inside and outside diameter surfaces and hydrogen pickup in the cladding were measured. The observed outer surface Zircaloy cladding corrosion oxide thicknesses of the test rods were similar to thicknesses measured for non-defected irradiation test rods. An analysis model gave results which were in reasonable agreement with the outer surface oxide thicknesses both of defected and non-defected rods. When the analysis procedure was modified to account for additional corrosion proportional to fission rate, the calculated values agreed well with measured inner oxide corrosion films values of defected rods. Hydrogen pickup in the defected rods was not directly proportional to local corrosion oxide weight gain as was the case for non-defected rods.

OSTI ID:
6980608
Report Number(s):
CONF-870314-
Country of Publication:
United States
Language:
English