MICROSTRUCTURAL INFLUENCE ON UNIFORM CORROSION OF ZIRCALOY NUCLEAR-FUEL CLADDINGS

Different chemical composition and production procedures were used to investigate the corrosion resistance of Zircaloy nuclear fuel cladding s in pressurized water reactors (PWR), The corrosion rate of the mater ials was measured for up to five years in the pressurized water reacto r Ringhals 3. Accelerated corrosion tests in autoclave were also perfo rmed in 400-degrees-C steam at 10.3 MPa. The microstructure of the mat erials was described with respect to intermetallic precipitate size di stribution and matrix composition between precipitates. According to a tom probe analysis, only about 10% (70-270 wppm) of the alloying eleme nts Fe, Cr and Ni added to the alloy remain in the matrix between prec ipitates. Different heat treatments and different corrosion resistance of the Zr-4 materials could not be related to significant differences in matrix composition.