Corrosion and hydrogen pickup of zirconium alloy fuel cladding in water cooled nuclear reactors are life-limiting phenomena for fuel. This thesis studies the fate of hydrogen liberated by waterside corrosion of Zircaloy-4 fuel cladding in Pressurized Water Reactors (PWRs): are the adsorbed protons incorporated into the oxide and eventually the metal, or are they evolved into molecular hydrogen and released into the coolant? Water chemistry modeling was used to understand effects of radiolysis and CRUD. Density functional theory (DFT) was used to investigate the role of oxidized Zr(Fe,Cr)2 second phase particles. Chemical potentials and the electron chemical potential were used to connect these two modeling efforts. A radiolysis model was developed for the primary loop of a PWR. Dose profiles accounting for fuel burnup, boron addition, axial power profiles, and a CRUD layer were produced. Dose rates to the bulk coolant increased by 21-22% with 12.5-75 pim thick CRUD layers. Radially-averaged core chemistry was compared to single-channel chemistry at individual fuel rods. Calculations showed that local chemistry was more oxidizing at high-power fuel and fuel with CRUD. Local hydrogen peroxide concentrations were up to 2.5 ppb higher than average levels of 5-8 ppb. Radiolysis results were used to compute chemical potentials and the corrosion potential. Marcus theory was applied to compare the band energies of oxides associated with Zircaloy-4 and the energy levels for proton reduction in PWR conditions. Hydrogen interactions with Cr203 and Fe203, both found in oxidized precipitates, were studied with DFT. Atomic adsorption of hydrogen was modeled on the Cr and Feterminated (0001) surfaces. Climbing Image-Nudged Elastic Band calculations were used to model the competing pathways of hydrogen migration into the subsurface and molecular hydrogen formation. A two-step mechanism for hydrogen recombination was identified consisting of: reduction of an adsorbed proton (H+) to a hydride ion (H-) and H2 formation from an adjacent adsorbed proton and hydride ion. Overall, results suggest that neither surface will be an easy entrance point for hydrogen ingress and that Cr203 is more likely to be involved in hydrogen evolution than the Fe203.

During normal operation in nuclear reactors, the nuclear fuel cladding corrodes as a result of exposure to high temperature cooling water. During this process, hydrogen can enter the zirconium-alloy of the fuel cladding, and under proper conditions, precipitate as brittle hydride platelets which can severely impact cladding ductility and fracture toughness. Hydrogen tends to migrate to and precipitate at colder spots. Because high local hydride concentrations increase the risk of cladding failure, it is important to predict the local hydrogen distribution. Hydrogen transport depends on different phenomena. Even though migration can only occur when the hydrogen is in solid solution, the cladding temperatures during operating condition allow a portion of the hydrogen to be in solid solution. Therefore, as hydrogen is picked up during the corrosion reaction between the cladding and the coolant, it can migrate following Ficks law and the Soret effect. Once the local hydrogen content reaches the terminal solid solubility for precipitation, the hydrogen will start precipitating as zirconium hydride. Previous work in our laboratory implemented a model that describes these different phenomena, into the 3D fuel performance code BISON. A first attempt to benchmark this model has been made in this study by comparing the results given by BISON to the hydrogen distribution measured in a nuclear fuel rod, which underwent a five cycle exposure at the Gravelines nuclear power plant. This was feasible because of the very detailed information about that reactor and fuel pin were available from the thesis of J.-H. Zhang in 1992. The benchmark performed with BISON showed very good agreement between the calculations experimental observation.The calculation above used a value of the precipitation kinetics parameter 2 based on a fit that ignored discrepancies in the early part of the precipitation process. The calculation was revised according to a new model. The new model, which assumes an initial dependence of the precipitation rate on (C_ss-TSS_p )^2 , provides a better fit for this date and we believe a more precise value of 2, which varies mildly with temperature.

LOCA simulation tests were performed in the QUENCH facility of KIT on fuel rod bundle scale. The first two tests using Zircaloy-4 claddings, out of a series of six tests with different cladding alloys, were already performed. The test conditions and results are described. The secondary hydrogenation of the Zircaloy-4 cladding tubes was investigated by means of neutron imaging. In the cladding of the inner rods, hydrogen enriched bended bands were found. They are non-symmetrical to the tube axis. The bands are located at the position where significant inner oxidation ends. X-ray diffractometry (XRD) measurements show that the hydrogen remains at least partially in the zirconium lattice. The formation of the hydrogen enriched bands results in an embrittlement of the cladding tubes and in changes of the fracture mode in tensile tests. The micro-hardness of the cladding increases at the band position. The reasons of the formation of these hydrogen enriched bands and the dependence of its appearance on the temperature scenario are discussed.