COBRA-TF evaluation and application For PWR steamline break DNB analysis

COBRA-TF (CTF) is a thermal hydraulic (T/H) subchannel code capable of calculating reversed flow, countercurrent flow and cross-flow using either three-dimensional (3D) Cartesian or subchannel coordinate formulations for fluid flow and heat transfer solutions. Recent software improvements on the code capability under the Consortium for Advanced Simulation of Light Water Reactors (CASL) program include software optimization, new closure models, and parallelization for modeling full reactor core T/H responses under PWR normal operating and accident conditions. Under the CASL development program, the CTF code is used for predicting the core T/H responses for both reactivity feedback and margin to departure from nucleate boiling (DNB) during a Pressurized Water Reactor (PWR) main steamline break (SLB) accident. An evaluation of the CTF code capability was performed for such application in comparison to experimental data from rod bundle tests simulating PWR fuel and SLB conditions. The CTF subchannel predictive capability was also compared with results from subchannel codes used in the industry. The CTF modeling capability was further evaluated using a full reactor core subchannel model to predict fluid and fuel conditions at the DNB-limiting time step of a PWR SLB case without offsite power on the Oak Ridge National Laboratory (ORNL) high-performance computing platform. Results of the capability evaluation indicate that CTF could potentially be applied to PWR accident analysis with respect to the DNB acceptance criterion.