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VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis. A 2x2 fuel assembly arrangement model was developed and simulated by VERA-CS, and uncertainty quantification and sensitivity analysis were performed with fourteen uncertain input parameters. The minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surface temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. Parameters used as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.

VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17x17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks' nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core retains design margin with respect to the MDNBR safety limit for both of the MSLB accident scenarios. The scenario with available offsite power was more restrictive in terms of MDNBR than the scenario without offsite power.

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