COBRA-TF parallelization and application to PWR reactor core subchannel DNB analysis

COBRA-TF (Coolant Boiling in Rod Arrays - Two Fluid) or CTF is a transient subchannel code, selected to be the reactor core thermal hydraulic (T/H) simulation tool in the multi-physics code, Virtual Environment for Reactor Applications Core Simulator (VERA-CS) development project of the Consortium for Advanced Simulation of Light water reactor (CASL) sponsored by the US Department of Energy. CTF has been improved and parallelized by CASL as part of its multi-physics software package to help the nuclear industry address operational and safety challenge problems, such as departure from nucleate boiling (DNB). In this paper, CTF's performance and capability are evaluated by modeling and analyzing full core 3D models of a 3- loop and a 4-loop pressurized water reactor (PWR) with all fuel assemblies modeled in subchannels. Calculations have been performed for DNB ratio (DNBR) predictions in complete loss-of-flow, low-flow main steam line break, and rod ejection at full power transients. Those applications demonstrate CTF's capabilities for modeling the entire reactor core in subchannels and simulating challenging PWR transients in preparation for coupled multi-physics analysis using the VERA-CS neutronic and T/H code system.