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A series of in-reactor corrosion tests was conducted in the G-7 loop of the Engineering Test Reactor (ETR) in pH 10 NH4OH at 270 to 280 C. Six zirconium alloys were exposed in as-etched and prefilmed (400 C steam) surface conditions. The six alloys were Zircaloy-2, Zircaloy-4, Zr-3Nb-1Sn, Zr-2.5Nb, Zr-1.2Cu-0.28Fe, and Zr-1.2Cr-0.08Fe. Exposures were 25, 37, 77, and 174 days in high-flux, low-flux, out-of-flux, and out-of-reactor environments. Results from the 174-day exposure are emphasized, but corrosion and hydriding data from the complete test series are compared.
A marked upturn in corrosion rate between the 77- and 174-day exposures is attributed to an increase in oxygen level near the end of a deoxygenating resin life. In the 174-day experiment, corrosion rates were accelerated on the six alloys at the high-flux position, but only on Zr-Cr-Fe specimens at the low-flux position. Prefilming adversely affected the corrosion of Zr-Cr-Fe, Zr-Cu-Fe, Zircaloy-2, and Zircaloy-4 specimens. The alloys containing niobium were the most resistant to in-flux corrosion and hybriding. Hydrogen-absorption rates were generally proportional to corrosion rates.
Monoclinic zirconium oxide was the principal phase on high-fluence (1.7 × 1021, n/cm2, > 1 MeV) specimens of the six alloys. Non-uniform oxidation was observed on Zr-Cr-Fe and Zr-2.5Nb alloys, including pustules, pitting, and cracks. Banded chromium segregation was observed in a Zr-Cr-Fe oxide and nodular copper segregation was observed in a Zr-Cu-Fe oxide; niobium distribution was relatively uniform in a Zr-2.5Nb specimen.
nuclear reactors, corrosion, hydrides, zirconium alloys, autoclaving, zirconium oxide, coolants
Johnson, A. B.
Senior Research Scientist, Chemistry Research Department, Pacific Northwest Laboratory, Battelle Memorial Institute, Richland, Wash.