Abstract: The gas-liquid two-phase flow widely exists in nuclear energy, chemical industry, petroleum engineering, and other industrial production fields. Two-phase flow will occur in pressurized water reactors during normal operation and accident conditions. The distribution of void fraction and interfacial area concentration in rod bundle channels has an important influence on the flow resistance, heat transfer, critical heat flux, and power distribution of the reactor. The complex geometric structure of rod bundle channels brings a great challenge to the local experimental measurement of two-phase flow. As a result, the existing database of the interfacial structure parameters in rod bundle channels is still insufficient, which greatly affects the accuracy of two-phase flow modeling. To clarify the transformation mechanism of the phase distribution in rod bundle subchannels and provide a database for future modeling, an experimental study of air-water two-phase flow in a 5×5 rod bundle channel was carried out in this paper. The four-sensor conductivity probe was used to measure the distribution of local two-phase flow parameters (void fraction, interfacial area concentration, bubble diameter, and bubble velocity) across the test section at 36.5 hydraulic diameters. The results show that with the increase of liquid phase velocity and the decrease of gas phase velocity, the “axial peak” gradually changes to the “wall peak” under the influence of radial force. The transition boundary of phase distribution is obtained and has a relatively accurate classification effect. The non-uniform void fraction distribution over the cross-section is observed, bubbles tend to gather in the central subchannels which are less affected by the wall and have a higher mass velocity. For the overall cross-section, the decrease of liquid phase velocity and the increase of gas phase velocity exacerbate the non-uniformity of bubble distribution. In the experimental conditions of this paper, the established correlation of void fraction and interfacial area concentration exhibits strong predictive capability. The average minimum relative error is ±18.2% and ±12.2% respectively. The drift flux model for the rod bundle channels should take into account various phase distribution types to enhance the modeling of distribution parameters. Furthermore, it is imperative to validate the model using comprehensive void fraction distribution databases to enhance the accuracy of prediction outcomes. Because of the small size of the bubble flow, the interfacial area concentration correlation established based on the vertical circular pipe database can also be used for rod bundle channels, but the modified correlation considering the influence of the bundle geometry on the bubble is more accurate. This study can provide a reference for the closure of the two-fluid model in the reactor thermal-hydraulic analysis program.