Hydrogen Embrittlement and Oxide Layer Effect in the Cathodically Charged Zircaloy-2.

Grzegorz Gajowiec,Marek Derezulko,Andrzej Zieliński,Beata Majkowska-Marzec,Bartosz Chmiela,Michał Bartmański

doi:10.3390/ma13081913

Grzegorz Gajowiec, Marek Derezulko + Show 4 more

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https://doi.org/10.3390/ma13081913

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Abstract

The present paper is aimed at determining the less investigated effects of hydrogen uptake on the microstructure and the mechanical behavior of the oxidized Zircaloy-2 alloy. The specimens were oxidized and charged with hydrogen. The different oxidation temperatures and cathodic current densities were applied. The scanning electron microscopy, X-ray electron diffraction spectroscopy, hydrogen absorption assessment, tensile, and nanoindentation tests were performed. At low oxidation temperatures, an appearance of numerous hydrides and cracks, and a slight change of mechanical properties were noticed. At high-temperature oxidation, the oxide layer prevented the hydrogen deterioration of the alloy. For nonoxidized samples, charged at different current density, nanoindentation tests showed that both hardness and Young’s modulus revealed the minims at specific current value and the stepwise decrease in hardness during hydrogen desorption. The obtained results are explained by the barrier effect of the oxide layer against hydrogen uptake, softening due to the interaction of hydrogen and dislocations nucleated by indentation test, and hardening caused by the decomposition of hydrides. The last phenomena may appear together and result in hydrogen embrittlement in forms of simultaneous hydrogen-enhanced localized plasticity and delayed hydride cracking.

Highlights

Some Zr alloys are used in the nuclear industry for fuel claddings [1,2,3], reflectors in light water reactors [4], and in spent nuclear fuel reprocessing plants [5,6]
The most severe nuclear event, the loss-of-coolant-accident (LOCA), goes through high-temperature oxidation of fuel pellets followed by rapid quenching because of an emergency coolant reflooding into the reactor core
Some short cracks appeared on surfaces oxidized at 700 ◦C, and they propagated and resulted in the descaling of the oxide layers obtained at 900 ◦C and 1000 ◦C

Summary

Introduction

Some Zr alloys are used in the nuclear industry for fuel claddings [1,2,3], reflectors in light water reactors [4], and in spent nuclear fuel reprocessing plants [5,6]. Zirconium alloys are applied to manufacture nuclear fuel pellets due to their low thermal neutron capture cross-section, proper strength properties, and excellent corrosion resistance in the cooling medium [7]. The nuclear fuel pellets are made of the Zr-Sn Zircaloys, Zr-Nb E110, E125, and E635, Zr-Sn-Nb Zirlo, Zr-Nb M5, and X5A alloys [7,8,9,10], Zr-Nb and Zr-Nb-Fe [11], and the 702 alloys [12,13]. The most important degradation mechanisms of zirconium alloys in the nuclear industry comprise high-temperature oxidation, delayed hydride cracking, electrochemical corrosion of waterside of fuel pellets, and pipelines, and creep at elevated temperatures [2,11]. The most severe nuclear event, the loss-of-coolant-accident (LOCA), goes through high-temperature oxidation of fuel pellets followed by rapid quenching because of an emergency coolant reflooding into the reactor core. The leakage of Materials 2020, 13, 1913; doi:10.3390/ma13081913 www.mdpi.com/journal/materials

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