12th International Symposium on Symbiotic Nuclear Power ( ISSNP-2026)

Accurate prediction of Departure from Nucleate Boiling (DNB) is crucial for the design and safety of Pressurized Water Reactors (PWRs). At present, standard subchannel codes are predominantly empirical and geometry-specific, leaving scope for the development of innovative techniques and approaches. This work presents a three-dimensional Eulerian two-fluid CFD model for the simulation and analysis of subcooled flow boiling and DNB in the 5×5 rod bundle of Test Section 74, which is representative of the peripheral region between two adjacent Combustion Engineering 14×14 fuel assemblies. The CFD framework represents seven unvaned spacer grids as porous regions calibrated against measured cold-flow loss coefficients. It employs RPI heat flux partitioning extended to the critical heat flux regime, together with a one-group Yao-Morel interfacial area transport equation and the SST k-ω turbulence model. Validation against Combustion Engineering test data showed agreement in grid loss coefficients within 5.9% and predicted single-phase rod surface temperatures within 2°F of the measurements, while under-predicting temperatures in the saturated regime by 3 to 7°F. At 2250 psia, the predicted critical heat flux exceeded the measured values by 15.0% and 15.6% at mass fluxes of 2.25 and 2.97 Mlb/hr·ft², respectively. These results demonstrate an economical and effective methodology for local DNB assessment in complex rod-bundle geometries.
 

CFD, Subcooled flow boiling, Departure from Nucleate Boiling, PWR rod bundle, Interfacial Area Transport Equation, Porous region