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Long-term static autoclave corrosion tests were conducted on Zircaloy-4 tube specimens in water at 633 K. The material variables included in this investigation were: annealing parameter range 10-17 to 10-19 h (with Q/R = 40 000 K). fabrication history variation of early and late beta-quenching steps, and final heat treatment variation from several levels of stress-relief-anneal to a recrystallization anneal. Specimens were weighed at intervals of approximately 28 days for a maximum corrosion test exposure of 1160 days. The weight gain data show transitions to an accelerated corrosion rate that became apparent at exposure times greater than 300 days. The transition times varied from 141 to 253 days. Metallographic and scanning electron microscopic examination showed that the metal-oxide interface had an irregular shape and the oxidation front appeared to progress into the metal by fracture of the hydride precipitates at the interface. Hydrogen absorption fractions were calculated for each specimen and were used to estimate the hydrogen level in each specimen at the transition point. The estimated hydrogen levels at transition agreed reasonably with the hydrogen solubility in Zircaloy at 633 K. The results indicate that the corrosion rate acceleration observed in autoclaves at long times is associated with the onset of hydride precipitation and subsequent hydride fracture at the metal-oxide interface. A review of in-reactor corrosion data from the literature reveals that a similar hydride associated corrosion rate acceleration occurs in low oxygen coolant conditions in PWRs and PHWRs. Hydride precipitation at the metal-oxide interface is the probable reason for the correlation between the time of long-term autoclave corrosion rate transition and the in-PWR cladding corrosion resistance. On the basis of the effect of hydrides on the in-reactor corrosion rate, it is suggested that a better ex-reactor corrosion test to simulate the in-PWR corrosion performance would be a water test at 633 K with an imposed heat flux. The effect of hydrides on the corrosion rate is strongly related to the size, distribution, and orientation of the hydrides in the Zircaloy cross section. An alloy development program is suggested to enhance the corrosion resistance of zirconium alloys in PWRs to extended burnups.
Zircaloys, zirconium alloys, low-oxygen coolant, autoclave corrosion, in-PWR corrosion, hydride precipitation, hydrogen uptake, hydride orientation, coherency, metal-oxide interface, embrittlement
ABB Combustion Engineering Nuclear Power, Windsor, CT