Oxidation of Zircaloy-4 during in situ proton irradiation and corrosion in PWR primary water
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The kinetics and morphology of oxides formed during in situ proton irradiation–corrosion experiment were analyzed. Experiments were conducted in 320 °C water with 3 wt ppm H2, while irradiated by a 3.2-MeV proton beam at a current density of 2 lA/cm2 producing a damage rate at 4.4 10 7 dpa/s. The resulting oxide was compared with reference samples corroded in an autoclave, and literature data found on in-reactor formed oxide. The corrosion rate of the sample irradiated in situ was 10 times faster than the in-pile corrosion rate. The cracked and porous irradiated oxide consisted of monoclinic equiaxed grains of zirconia with a preferential orientation of the oxide grains. Second phase particles (SPPs) consumed by the oxidation front were rapidly oxidized, but no SPPs were amorphized or dissolved in the metal matrix of the irradiated sample.
I. INTRODUCTION
Zirconium alloys have been used in the nuclear industry for decades, as cladding materials, due to their superior corrosion resistance and low absorption cross-section for thermal neutrons. Current trends toward extending burnup, increased outlet temperatures and plant life extension in nuclear power facilities require a better mechanistic understanding of the corrosion under irradiation. The life-limiting factors for Zircaloy-4 in a reactor are normally corrosion and embrittlement caused by hydrogen absorption. Since essentially, all of the hydrogen absorbed into the metal was created during the corrosion process, it is crucial to understand the role of radiation-enhanced corrosion, hence, its impact on radiation-enhanced hydrogen absorption. Studies based on in-pile and out-of-pile weight gain data suggest that neutron radiation has a large impact on the corrosion rate of zirconium alloys during the service.1,2 Garzarolli et al.,1 showed a comparison of the in-pile pressurized water reactor (PWR) data with the out-of-pile behavior of Zircaloy-4 in deionized water, which indicates the same corrosion behavior up to a weight gain of about 50 mg/dm2, which is the equivalent of 3.4 lm of oxide (14.7 mg/dm2 1 lm oxide3), after which it is enhanced by irradiation. At higher fluences, the corrosion rate is increased by a factor of five, indicating that radiation had an increase in corrosion rate under reactor operating conditions.
Contributing Editor: Joel Ribis a) Address all correspondence to this author. e-mail: [email protected] DOI: 10.1557/jmr.2014.408 J. Mater. Res., Vol. 30, No. 9, May 14, 2015
Asher et al.4 found that zirconium irradiated in-reactor in moist carbon dioxide–air mixtures had oxygen weight gains of more than five times that in the unirradiated state. Bradhurst et al.5 observed in-reactor corrosion rates that were ten times greater than those conducted out-of-reactor and attributed part of the difference to greater permeability of the oxide irradiated in-reactor. Furthermore, old data on in-reactor exposure of Zircaloy-2 show an increase in oxide weight gain of 40-fold and a strong linear dependence with neutron flux.6 These data suggest that displ
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