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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EXECUTIVE SUMMARY Under the Reactor Product Line (RPL) of DOE/NE's Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, an SFR System Analysis Module is being developed at Argonne National Laboratory for whole-plant safety analysis. This tool will simulate tightly coupled physical phenomena-including nuclear fission, heat transfer, fluid dynamics, and thermal-mechanical response-in SFR structures, systems, and components. It is based on the MOOSE (Multi-physics Object-Oriented Simulation Environment) framework, which relies upon open-source libraries such as libMesh and PETSc for mesh generation, finite element analysis, and numerical solutions. This development is a coordinated effort along with the development of RELAP-7, which is an advanced safety analysis tool for light-water reactors developed at Idaho National Laboratory. The SFR Module is aimed to model and simulate the SFR systems with much higher fidelity and with well-defined and validated prediction capabilities. It will provide fast-running, modest-fidelity, whole-plant transient analyses capability, which is essential for fast turnaround design scoping and engineering analyses. The SFR System Module is being built based on RELAP-7 and the MOOSE framework. It leverages the common features between LWRs and SFRs (e.g., single-phase flow in a pipe and steam-system modeling for the balance of plant). The existing RELAP-7 physics models and component library are directly available for use in the SFR System Module. On the other hand, the SFR Module effort assists with RELAP-7 development by providing code verification and contributing general physics models and components applicable to all reactor types. Although the MOOSE and RELAP-7 based SFR system analysis module is a relative new effort (less than two years), significant accomplishments have been achieved. A 1-D FEM flow model using a pressure-based formulation with numerical stabilization schemes has been developed for use in incompressible sodium flows. A set of SFR-specific physics models and component has also been developed. The SFR primary system simulation capabilities of the SFR Module have been demonstrated by simulating the early stage of the Protected Loss-Of-Flow (PLOF) accident in the Advanced Burner Test Reactor (ABTR). Both the steady-state and PLOF transient simulation results are compared with the SAS4A/SASSYS-1 simulation results. It is confirmed that major physics phenomena in the primary coolant loop during the transient can be captured by the SFR Module. The SFR Module also emphasizes providing multi-scale multi-physics modeling capabilities by integrating with other higher-fidelity advanced simulation tools. This investigation is important to the integration between the MOOSE-based system code and the high-fidelity and medium-fidelity advanced simulation capabilities developed under the NEAMS RPL. The multi-scale coupling capability has been demonstrated in the coupled SFR Module and STAR-CCM+ code simulation of the ABTR PLOF transient. The importance of the multi-re...

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