Hydrogen uptake in zirconium alloys is a major limiting factor of nuclear fuel cladding as subsequent hydride precipitation can lead to potential failure via delayed cracking (DHC). The specific mechanisms DHC have been focus earlier research, however there still no clear definition the resulting crystallography and sizes at various temperatures crack propagation. In this work, novel thermo-mec...

Abstract Nuclear fuel performance modeling and simulation are critical tasks for nuclear design optimization safety analysis under normal transient conditions. Fuel is a complicated phenomenon that involves thermal, mechanical, irradiation mechanisms requires special multiphysics modules. In this study, model was developed using the COMSOL Multiphysics platform. The performed 2D axis-symmetric ...

Inner-side coatings have been proposed as a complementary solution within the accident tolerant fuel (ATF) framework, to provide enhanced protection for nuclear cladding. Unlike external surface, degradation of irradiated internal cladding surface has not studied extensively. Fission fragments produced during fission is one key players in this degradation. This study aimed estimate minimum thic...

The aim of this study is to investigate the potential improvement accident-tolerant fuels in pressurized water reactors for replacing existing reference zircaloy (Zr) fuel-cladding systems. Three main strategies improving are investigated: enhancement present state-of-the-art zirconium system improve oxidation resistance, replacement current referenced material with an alternative high-performa...

The temperature of zirconium alloy cladding on the postulated spent nuclear fuel pool complete loss coolant accident is abruptly increased at a certain time and almost fully oxidized to weak ZrO2 in air. This abrupt escalation phenomenon induced by air-oxidation breakaway called fire. Although an kinetic model correlated between has been implemented MELCOR code, it likely bring about unexpected...

Uranium dioxide (UO2) is the major nuclear fuel component of fission power reactors. A key concern during severe accidents is the melting and leakage of radioactive UO2 as it corrodes through its zirconium cladding and steel containment. Yet, the very high temperatures (>3140 kelvin) and chemical reactivity of molten UO2 have prevented structural studies. In this work, we combine laser heating,...

Modeling fuel behavior requires an accurate description of the cladding stress response for both operational and safety considerations. The transient creep response of Zirconium alloys is commonly modeled using a strain hardening rule which is known to hold in cases with monotonously increasing stresses. However, the strain hardening rule is experimentally known to fail in scenarios such as loa...