Release and speciation of carbon from Zircaloy-4 in anaerobic and highly alkaline conditions: Comparison of simple immersion and potentiostatic corrosion tests

Sebastien Caes,Frank Druyts,Peter Thomas

doi:10.1017/rdc.2018.105

Abstract

ABSTRACTThe gas release and speciation of carbon species from irradiated and unirradiated Zircaloy-4 samples, representative for the fuel cladding as used in Belgian nuclear power plants, were studied in a saturated Ca(OH)2 solution in anaerobic conditions. This environment is relevant for the Belgian Supercontainer design, as perceived for the geological disposal of high-level nuclear waste. To achieve this, we performed simple immersion and potentiostatic corrosion tests. Potentiodynamic polarization curves, recorded prior to the potentiostatic tests, revealed that irradiation seems to induce changes on the Zircaloy-4 corrosion behavior, such as a shift of the corrosion potential. Potentiostatic corrosion tests on unirradiated Zircaloy-4 provided a corrosion rate of ~54 nm/yr over a 7 day-experiment, whilst a corrosion rate of only ~4 nm/yr was calculated for the irradiated sample. Gas chromatography revealed that during simple immersion tests, which lasted 195 days, hydrogen, methane, ethane, and CO2 were produced, with methane being the major compound. Assuming that all carbon released from the metal was transformed into gaseous compounds, this yields to a corrosion rate ranging from 57 to 84 nm/yr for the irradiated sample. However, caution has to be taken on these corrosion rate and more tests should be performed to confirm these results.

Highlights

Zirconium alloys are widely used in nuclear reactors as fuel cladding because they offer a low neutron cross section, reasonable mechanical properties, and adequate corrosion resistance in high-temperature water (Garzarolli et al 1996)
The aim of this work was to investigate the release of carbon, and carbon-14, from Zircaloy-4, representative for the fuel claddings as used in Belgian nuclear power plants, and the carbon speciation in a cementitious and anaerobic environment, which is relevant for the Belgian Supercontainer design, as perceived for the geological disposal of high-level waste (Bel et al 2006)
This work investigated the release of carbon from Zircaloy-4 representative for the fuel cladding as used in the Belgian nuclear power plants, and the carbon speciation in a cementitious environment, which is relevant for the Belgian Supercontainer design, as perceived for the geological disposal of high-level waste

Summary

Introduction

Zirconium alloys are widely used in nuclear reactors as fuel cladding because they offer a low neutron cross section, reasonable mechanical properties, and adequate corrosion resistance in high-temperature water (Garzarolli et al 1996). % Cr, was the only zirconium alloy used in Belgian reactors. Zirconium alloys possess a high corrosion resistance to uniform and localized corrosion due to the formation of a zirconium dioxide protective layer (Lefebvre and Lemaignan 1997; Mogoda 1999; Hillner et al 2000; Motta et al 2015). The corrosion kinetics follow a cubic law. When the oxide layer reaches a thickness of 2–3 μm, a series of successive cubic curves often approximated by a linear law are observed (Hillner et al 2000)

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