Published: Jan 1969
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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.
Paper ID: STP43835S