Long-Term Oxidation of Zirconium Alloy in Simulated Nuclear Reactor Primary Coolant—Experiments and Modeling

Author:

Betova Iva1,Bojinov Martin2ORCID,Karastoyanov Vasil2

Affiliation:

1. Institute of Electrochemistry and Energy Systems, Bulgarian Academy of Sciences, 1113 Sofia, Bulgaria

2. Department of Physical Chemistry, University of Chemical Technology and Metallurgy, 1756 Sofia, Bulgaria

Abstract

Oxidation of Zr-1%Nb fuel cladding alloy in simulated primary coolant of a pressurized water nuclear reactor is followed by in-situ electrochemical impedance spectroscopy. In-depth composition and thickness of the oxide are estimated by ex-situ analytical techniques. A kinetic model of the oxidation process featuring interfacial reactions of metal oxidation and water reduction, as well as electron and ion transport through the oxide governed by diffusion-migration, is parameterized by quantitative comparison to impedance data. The effects of compressive stress on diffusion and ionic space charge on migration of ionic point defects are introduced to rationalize the dependence of transport parameters on thickness (or oxidation time). The influence of ex-situ and in-situ hydrogen charging on kinetic and transport parameters is also studied.

Funder

National Scientific Fund of Bulgaria

Publisher

MDPI AG

Subject

General Materials Science

Reference32 articles.

1. Tupin, M. (2020). Nuclear Corrosion, European Federation of Corrosion, Elsevier.

2. Yagnik, S., and Garde, A. (2019). Structural Alloys for Nuclear Energy Applications, Elsevier.

3. Fromhold, A.T. (1976). Theory of Metal Oxidation-Vol. I: Fundamentals, North-Holland.

4. Some thoughts on the mechanisms of in-reactor corrosion of zirconium alloys;Cox;J. Nucl. Mater.,2005

5. Comstock, R.J., and Motta, A.T. (2018). Zirconium in the Nuclear Industry, Proceedings of the 18th International Symposium, ASTM International.

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