IMPLEMENTATION AND VALIDATION OF A FISSION GAS RELEASE MODEL FOR CTFFUEL USING THE NEA/OECD IFPE DATABASE

The nuclear reactors themselves are complex systems whose responses driven by interactions between different physics phenomena within the reactor core. Traditionally, have been analyzed separately and its interaction considered via boundary conditions or closure models. However, in parallel with development of computational technology, multi-physics coupled simulations being used to obtain accurate predictions thanks consideration feedback effects on fly (on-line). In fuel temperature is an important parameter cross sections at given neutron kinetics codes. An prediction profile rod involve several such as kinetics, mechanics, material behavior properties, heat transfer, thermal-hydraulics, even chemistry. pellet clad gap conductance possibly most source uncertainty solution conductivity equation prediction. depends two effects: distance gas species that fill gap. this research work, authors focused improving CTFFuel implementing a fission release model code. objective contribution implementation transient validation using experimental data available OECD/NEA International Fuel Performance Experiments (IFPE) database. isolated transfer capability framework CTF code, state-of-the-art version Coolant Boiling Rod Arrays Code – Two-Fluid (COBRA-TF) sub-channel thermal-hydraulic code jointly developed North Carolina State University (NCSU) Oak Ridge National Laboratory (ORNL) US Department Energy (DOE) Consortium for Advanced Simulation LWRs (CASL).