In-reactor and long-term autoclave corrosion data have been obtained on ZIRLO and three variants of Zircaloy-4: conventional (1.5% tin), low-tin, and beta-treated. In-reactor data from demonstration assemblies irradiated in the Virginia Power Company's North Anna Unit 1 reactor demonstrate the superiority of ZIRLO and, to a lesser extent, low-tin Zircaloy-4 over conventional Zircaloy-4. After two cycles of irradiation to an assembly burnup of 37.8 GWD/MTU, the average axial peak corrosion of ZIRLO was 32% that of conventional Zircaloy-4. Low-tin and beta-treated materials displayed average peak oxides 76% and 150% of that formed on conventional Zircaloy-4, respectively.

Autoclave corrosion tests of archive tubing have been performed in 633 K water, 672 K, 700 K, and 727 K steam, and in 633 K water containing 70 and 210 ppm lithium as the hydroxide. Correlation of the in-reactor data with the autoclave data indicates that the 633 K pure water test is the best qualitative indicator of in-reactor corrosion performance, and the 672 K steam test the poorest. Differences in in-reactor corrosion between ZIRLO and the Zircaloy-4 materials are consistent with the relative behavior of these materials in lithium hydroxide solutions. The relationships among the in-reactor and autoclave corrosion data, the microstructures, and the processing are discussed.

In addition to improved corrosion resistance, ZIRLO exhibits improved dimensional stability over Zircaloy-4. The in-reactor creep of ZIRLO is confirmed to be about 80% of that of Zircaloy-4, and irradiation growth is observed to be about 50% that of Zircaloy-4. These data are also presented and discussed.