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.

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

“…Recently, work performed under the CASL program (Y. Sung, et al 2015b) (Kucukboyaci, et al 2016) used VERA-CS to assess the core response to a MSLB event that poses a challenge to the departure from nucleate boiling (DNB) criteria. Since the minimum DNB ratio (MDNBR) occurs minutes after initiation of the transient, the reactor core response at the MDNBR time step can be simulated in a quasi-steadystate mode similar to the current industrial practice (Sherder and McHugh 1998).…”

“…The previous work under the CASL program (Y. Sung, et al 2015a) (Kucukboyaci, et al 2016) considered one HF and one LF scenario with estimated core inlet distributions. Based on the previous work, the current BEPU analysis validates the conclusion of the DNB limiting case by accounting for uncertainties in the core boundary conditions and core modeling parameters and assessing their effects on the FOM.…”

“…The reactor core inlet temperature and mass flux distributions on an assembly-by-assembly basis from the CFD simulations as input to VERA-CS for the high and low flow cases are shown in Figure 3 through Figure 6. A VERA-CS sensitivity study (Kucukboyaci, et al 2016) concluded that the DNBR responses under the MSLB conditions were insensitive to uncertainties in the inlet temperature and flow distributions from the CFD predictions. VERA-CS has the capability to define inlet temperature ( Figure 3 and Figure 4) and mass flux ( Figure 5 and Figure 6) distributions on an assembly-by-assembly basis and that capability was used in the analysis.…”

VERA Integration in the NEAMS Workbench

“…The Virtual Environment for Reactor Applications (VERA) [1] is a state-of-the-art fully coupled, pinresolved reactor core simulator. In addition to high-fidelity core simulation for typical reactor depletion and transient calculations [2], VERA also enables more advanced analyses such as calculations of vessel fluence [3], departure from nucleate boiling (DNB) [4,5], fuel performance [6], and various multiphysics phenomena such as pellet-clad Interaction (PCI) [7] and crud-induced power shift (CIPS) [8]. Although significant effort has been made to ensure the usability of VERA, constructing and validating reactor models, significant remaining work includes ensuring successful execution of the reactor models, as well as subsequent post-processing of results.…”