Current Pressurized Water Reactors (PWR) fuel assembly thermal-hydraulic (T/H) analyses are performed on a subchannel basis that neglects detailed heat transfer and flow distributions surrounding fuel rods. Subchannel codes such as VIPREW require input of thermal mixing and hydraulic loss coefficients that are obtained from costly experiments. Fuel thermal margin or performance is quantified in terms of Departure from Nuclear Boiling Ratio (DNBR) for PWR applications or Critical Power Ratio (CPR) for Boiling Water Reactors. DNBR and CPR predictions for reactor design and safety analysis rely on empirical correlations that are developed and qualified from costly rod bundle water DNB tests. Demands for extended power uprate, high fuel burnup, zero fuel failure, and new nuclear plant designs require a revolutionary advancement in T/H capability for better understanding of coolant behavior and more accurate predictions of thermal margin of the Light Water Reactor (LWR) core and fuel designs under normal operation and postulated accident conditions. Computational Fluid Dynamics (CFD) has been used in many aspects of PWR fuel designs in Westinghouse. Significant advancement in 2-phase flow modeling has been made in the recent years. This paper will illustrate CFD–based DNB modeling development in Westinghouse Nuclear Fuel. A 5×5 test bundle PWR experiment from the ODEN DNB test facility was modeled in CFD using a relatively new 2-phase boiling model. The model geometry included the details of the mixing vane spacer grids. When compared to the test data, the CFD model demonstrated that the DNB power was reasonably predicted. The CFD model also revealed the detailed flow behavior and the 2-phase flow distribution, both of which will be beneficial for the development of new grid spacers.

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