Investigation of Hydrogen Behaviour in Irradiated Zirconium-Based Fuel Claddings by Neutron Imaging and Complementary Techniques

Zirconium-based cladding tubes, mainly used as fuel cladding materials in light water reactors, serve as the primary barrier insulating the radioactive fission products from the environment. During operation, the cladding interacts with the coolant water, leading to oxidation that produces hydrogen as a byproduct. Hydrogen is either released into the reactor water, trapped in the zirconium oxide or can diffuse into the metal. When the hydrogen concentration in the metal exceeds the solid solubility limit, brittle zirconium hydrides form, negatively affecting the cladding's mechanical properties and potentially compromising its integrity under certain conditions. Thus, studying hydrogen embrittlement in zirconium alloys is crucial for ensuring the integrity of nuclear fuel claddings throughout their lifecycle. Particularly, understanding the distribution of hydrogen and hydrides under dry storage conditions, where the risk of hydrogen embrittlement is higher, is a critical research area for regulators, manufacturers, utilities, and academia. While much research has been performed on the concentration and temperature-driven hydrogen diffusion and hydride precipitation with non-irradiated claddings, investigations with reactor irradiated claddings are scarce. This PhD thesis explores hydrogen and hydrides distribution behaviour in both non-irradiated and irradiated Zr-based claddings, including their oxide layers, focusing on Zircaloy-4 (single material cladding) and DXD4 (duplex material cladding) used in Swiss pressurized water reactors. High-resolution neutron imaging (HR-NI) procedure was further developed for this purpose, particularly for irradiated samples, and served as the primary analytical tool. The thesis comprises two main parts. The first part quantitatively characterizes hydrogen redistribution after various thermal and thermo-mechanical test conditions and discusses the mechanisms influencing both single and duplex material claddings. Comparisons are made between irradiated samples with five reactor cycles and non-irradiated test samples. The second part investigates trapped hydrogen in the zirconium oxide layer and its outgassing. Findings indicate that in single material claddings, hydrogen redistribution and hydride formation are influenced by stress gradients, with irradiated materials exhibiting behaviours similar to non-irradiated materials. Duplex material claddings show a complex interplay of diffusion and precipitation mechanisms. A dense hydride accumulation zone in the liner forms during slow cooling, with concentrations varying from 1000 to 2500 wppm depending on the test conditions. Irradiated materials demonstrate slightly varied behaviors due to multiple influencing factors. Using ptychographic X-ray computed tomography (PXCT), the bulk mass density of irradiated zirconium oxide was measured to be about 5.11 ± 0.80 g/cm³. A higher hydrogen concentration was detected in irradiated zirconium oxide than in the metal, and