Evaluation of a CeO2 candidate material for nuclear experimental applications through thermal-hydraulic analysis with COBRA-TF in VVER-1200 reactor

Uranium dioxide (UO₂) has long been used as the standard fuel in pressurized water reactors (PWRs) such as the VVER-1200. However, its radioactive nature, high fabrication cost, and regulatory constraints present significant challenges for experimental studies. In this work, cerium dioxide (CeO₂), a non-fissile and chemically stable ceramic, is investigated as a potential surrogate for UO₂ in thermal-hydraulic experiments. Using the COBRA-TF (CTF) subchannel analysis code, a VVER-1200 fuel assembly was modeled to simulate the thermal behavior of both UO₂ and CeO₂ under steady-state and transient conditions, including reactivity insertion and loss-of-flow accidents.Simulation results indicate that CeO₂’s lower thermal conductivity and specific heat capacity lead to higher fuel centerline temperatures—up to 200 ℃ higher during normal operation and approximately 120 ℃ during transient events. While these elevated temperatures reflect conservative predictions, the fuel surface and cladding temperatures remain comparable to UO₂, supporting CeO₂'s suitability for experimental investigations such as cladding oxidation or material testing. Additionally, similar Critical Heat Flux Ratio (CHFR) values in both materials confirm CeO₂'s applicability in replicating coolant thermal conditions. These findings suggest that CeO₂ can serve as a viable experimental substitute to study key fuel and cladding behaviors without the complexities of handling radioactive UO₂.