Abstract: Subcooled boiling heat transfer is highly concerned in pressurized water reactor (PWR) research and design (R&D). Void fraction is an important parameter for neutronics/thermal-hydraulic (T-H) coupling analysis and a dominant parameter of departure from nucleate boiling (DNB). In this study, computational fluid dynamics (CFD) method was adopted to predict the void fraction in fuel assembly under subcooled boiling condition to verify the capability of two-phase flow CFD model under a wide T-H parameter range, which is a critical issue in the T-H R&D of PWR fuel assembly. The two-phase flow CFD model was based on the two-fluid model framework. Wall heat flux partition model with modified sub-models was used to predict subcooled boiling and void generation. Interface force and bubble breakup/coalescence models were considered. The 5×5 PSBT (OECD/NRC benchmark based on NUPEC PWR subchannel and bundle tests) fuel assembly rod bundle was simulated and its averaged void fraction measurement data of the central four subchannels were used for comparison. The B5 test serial was setup with a 5×5 full length scale bundle with uniform-axial power distribution (uniform-APD). The range of pressure, mass flux, inlet subcooling and local thermal equilibrium quality is 4.79-16.59 MPa, 555-4194 kg/(m²·s), 17.5-117.2 K and -0.15-0.14, respectively. The comparison results show that there are 51% prediction data within the measurement uncertainty (±0.04) and 88% prediction data within 2 times of measurement uncertainty (±0.08). By implementing the two-phase flow CFD model in a wide range of T-H parameters, the prediction and measurement value are good agreement, which shows the capability of CFD method and its potential for further application.