Specific Aspects of Internal Corrosion of Nuclear Clad Made of Zircaloy

TL;DR

This paper investigates the internal corrosion of Zircaloy cladding in nuclear reactors, highlighting its unique behavior, stress effects, and differences from external corrosion through simulations and analysis.

Contribution

It provides new insights into the specific internal corrosion mechanisms of Zircaloy cladding and their impact on stress distribution and oxygen diffusion in nuclear fuel rods.

Findings

Internal corrosion forms a 5-15 μm oxide layer on Zircaloy.

Stress fields influence oxygen diffusion profiles.

Differences between inner and outer corrosion affect diffusion phenomena.

Abstract

In PWR, the Zircaloy based clad is the first safety barrier of the fuel rod, it must prevent the dispersion of the radioactive elements, which are formed by fission inside the UO2 pellets filling the clad. We focus here on internal corrosion that occurs when the clad is in tight contact with the UO2 pellet. In this situation, with temperature of 400^{\circ}C on the internal surface of the clad, a layer of oxidised Zircaloy is formed with a thickness ranging from 5 to 15 $\mu$m. In this paper, we will underline the specific behaviour of this internal corrosion layer compared to wet corrosion of Zircaloy. Simulations will underline the differences of stress field and their influences on corresponding dissolved oxygen profiles. The reasons for these differences will be discussed as function of the mechanical state at inner surface of the clad which is highly compressed. Differences between mechanical conditions generated by an inner or outer corrosion of the clad are studied and their influences on the diffusion phenomena are highlighted.