VERA-CS Modeling and Simulation of PWR Main Steam Line Break Core Response to DNB

Abstract

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics (T/H), and fuel temperature components with an isotopic depletion capability. The neutronics capability is based on the Michigan Parallel Characteristics Transport Code (MPACT), a three-dimensional whole-core transport code. The T/H and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to the departure from nucleate boiling (DNB) ratio at the most limiting point of a postulated pressurized water reactor main steam line break event initiated at the hot zero power, either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power, where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady-state reactor core response under the main steam line break accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.