The influence of nuclear radiation on zirconium alloy corrosion was studied in the North (1D) loop of the Advanced Test Reactor (ATR). The loop was new and therefore was free of fissionable contamination. The principal corrosion parameters were: alloy composition (Zircaloy-2, Zircaloy-4, Zr-2.5Nb, and crystal bar zirconium); flux level (unirradiated, 6 × 1012 to 1.7 × 1014 neutrons (n)/cm2 s > 1 MeV); metallurgical condition (Zircaloy-4 and Zr-2.5Nb in two conditions); oxygen concentration (0.6 ppm O2 and < 0.05 ppm O2); and surface pretreatment (as-etched, thin prefilms, and thick prefilms). Maximum exposures were 61.6d (oxygenated) and 102.1 d (deoxygenated) at 280°C, 2000 psig, in pH 10 LiOH. In the oxygenated test, corrosion rates were substantially accelerated at high-flux locations. At the low-flux location, corrosion rates were unaffected or slightly accelerated on Zircaloy-2 and -4, substantially accelerated on zirconium, and slightly reduced on Zr-2.5Nb. Thick prefilms tended to suppress accelerated corrosion in the oxygenated system. In the deoxygenated test, corrosion rates were substantially accelerated for zirconium, but were only slightly above unirradiated values for the other materials. The results from the ATR loop paralleled previous results from the ETR G-7 loop, indicating that dissolved oxygen, rather than fissionable contamination, was the principal cause of accelerated corrosion in both systems. The Zr-2.5Nb alloy (quenched, cold-worked, aged) had the lowest weight gains and hydrogen absorption in both oxygenated and deoxygenated tests, but showed some pustule formation in the oxygenated test. Zircaloy-2 coupons which had oxidized rapidly to 120 to 160 mg/dm2 (2.0 to 2.6 mg/dm2d) in the oxygenated system were transferred to the deoxygenated system. After transfer the corrosion rates decreased to 0.5 to 0.6 mg/dm2d. Inspection of coupons under polarized light revealed evidence of oxide porosity on coupons undergoing accelerated corrosion.
The heat-treatable 2.5Nb zirconium alloy is of interest to nuclear-reactor designers because it has greater strength than the Zircaloys. Its corrosion behavior is affected by its metallurgical structure, the chemical composition of the environment, and the presence of reactor radiation. Desirable mechanical properties are obtained by quenching from a temperature high in the (α + β)-phase region (generally 850 to 880 C) and stabilizing the structure by aging in the α-phase region (500 to 575 C). Cold work between quenching and aging enhances some properties, notably corrosion resistance. This is illustrated by results of corrosion experiments in aqueous and gaseous media. The corrosion resistance after different amounts of cold work and aging may be correlated with the sub-microstructure developed, as shown by electron microscopy. Evidence from several independent experiments shows that the alloy is much less sensitive to radiation enhancement of corrosion than is Zircaloy, even under strongly oxidizing conditions, such as neutral boiling water. In many cases the oxidation in-flux is less than that out-of-flux in the same coolant stream. The 2.5Nb zirconium alloy has found nuclear application as pressure tubes for heavy-water moderated reactors, both with pressurized water and boiling water coolant, and as fuel cladding in organic coolant. Results of irradiation experiments suggest it would be useful as fuel cladding in boiling-water-cooled reactors and for conditions which exceed the upper temperature limit of Zircaloy cladding.
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