The integrity of structural components is a major concern in reactor safety. Cladding materials can fail when hydrogen uptake is in excess of the terminal solid solubility (TSS). Previous studies on hydrogenation or secondary hydriding (SH) focused mainly on loss of coolant accident (LOCA) and fuel ballooning. Most of these studies are carried out in autoclave conditions and they preclude the contribution of ionizing radiations (neutrons and gamma) to hydrogen generation during normal reactor condition. This study is devoted to the hydrogenation of Zircaloy-4 (Zry-4) cladding material during normal reactor operation and during the early phase of core degradation in a LOCA event. Radiolysis and corrosion are considered as hydrogen sources during normal reactor operation while corrosion is considered in the early phase of core degradation. Using appropriate initial and boundary conditions peculiar to pressurized water reactors (PWRs), diffusion equation is solved in a defective Zry-4 fuel system using COMSOL Multiphysics 5.2. Results show that the average corrosive source term (1.279E−4 mol m−3 s−1) is higher than the radiolytic source term (3.6594E−7 mol m−3 s−1) by a factor of 350 during normal reactor operation. By integrating in two kinetics regimes and setting 6 years as the maximum lifetime of Zry-4, the amount of H dissolved in the cladding material at a temperature of 633 K are 6.54E−03 and 2.02 wt. ppm due to radiolysis and corrosion respectively. These values are safe compared to the TSS of dissolution (CTSSD) of Zry-4 which is 117 wt. ppm at the same referenced temperature. However, in the early phase of core degradation where the source term is 9.405E−01 mol m−3 s−1, SH is observed in 13 days. A comparison is made between hydrogenation by integration in two regimes and hydrogenation using Sievert’s law. The results obtained in this study are useful for the prediction of the onset of hydrogen embrittlement, and to determine the service time of Zry-4 cladding material.

To optimize fission fuel and protect cladding integrity, this work investigates shadow corrosion in a one-fourth circular electrode geometry. The anodic corrosion of Zircaloy-2 (Zry-2) was investigated in a circular geometry electrode configuration under reactor operating conditions. The impact of gamma and neutron radiations on water conductivity and shadow corrosion was examined under two different cathodes. This work also investigates the effect of current exchange density and the cathodic Tafel coefficient on the cathodic current. Using COMSOL Multiphysics 5.2, the Laplace equation was solved to obtain the electrostatic potential and current density distributions in the studied domain. When the distance d between the anode (Zry-2) and cathode (platinum/nickel) is ≤0.5 mm, it was observed that a uniform oxide layer of thickness 20 µm grew on the smooth internal surface of Zry-2 for corrosion lasting 1166 h. When d > 0.5 mm, the oxide thickness falls in a manner dictated by the degree of dissociation α of the electrolyte. At a cladding gap of 0.08 mm, a radiation-enhanced uniform corrosion rate of 2.405 10−1 mmpy was obtained for Zry-2. This value is 142 times greater than that obtained at room temperature in the absence of radiation. It was also observed that the corrosion rate falls at higher cladding gaps, and the rate of change depends on the degree of dissociation. Other phenomena such as the dynamics of shadow corrosion under varying electrode separation and electrolyte conductivities, as well as extensive evaluation of critical fuel cladding parameters, are presented in this work.

One of the most common threats to the integrity of reactor fuel cladding is the geometric imperfections such as the missing pellet surface (MPS) that produces a remarkable surge in the local fuel-clad gap. The cooling water could occupy this gap leading to secondary hydriding (SH) and hydrogen embrittlement. Most studies on this subject have identified extensive radiolysis in boiling water reactors (BWRs) to be responsible for hydrogen evolution during accident conditions. However, the quantitative determination of hydrogen and how it affects zircaloy-2 during normal reactor operation has not been given adequate attention. To bridge this gap and to better predict the onset of cladding failure, this study investigates secondary hydriding and its characteristic phenomenon in a fuel cladded with zircaloy-2. Multiphysics model is used to model diffusion of heat and hydrogen, then the effect of an intermediary porous/non-protective oxide layer and the impact of dose rate from different types of radiation are studied. The contributions of the source term due to radiolysis of neutrons (n), gamma (γ), alpha (α) and beta (β) radiations are also considered. Results showed that neutrons, having a maximum dose rate of 39.9 kGy/s accounts for over 99% of ZrH precipitation in a UO2 fuel with MPS. The effect of oxide (ZrO2) thickness in SH, and the derivative of oxide thickness are also discussed.

The underline challenge of low thermal conductivity which is responsible for high centerline temperature in Uranium-dioxide (UO2) nuclear plant fuel type associated with the current generation of commercial reactors remains a huge concern to the nuclear power industry. Although, researchers in the nuclear industry have proposed uranium mononitride (UN) as a promising candidate for accident tolerant fuel and Generation-IV nuclear reactor fuels, unfortunately, there is still a lack of clear understanding of the point defects influence on the thermal conductivity of UN. This paper applies the reverse-NEMD method to study the effect of point defects on phonon thermal conductivity in UN perfect crystal. Experimental and simulation relevant literatures were reviewed to analyze the influence of defects on the phonon thermal conductivity of UN. The thermal conductivity of uranium mononitride increases quickly with the temperature and the phonon contribution decreases with an increase in vacancies concentration. Uranium vacancy has a detrimental effect on the phonon thermal conductivity of UN at lower vacancy concentration. Results from this research would help with the understanding and application of UN in nuclear engineering.