Abstract: Direct contact condensation is a highly efficient and compact heat transfer technology, widely applied in boiling water reactors and small modular reactors for nuclear reactor safety, especially in suppression containments and emergency coolant systems. During accident scenarios like loss-of-coolant accidents (LOCAs), high-temperature and high-pressure steam mixed with non-condensable gases is rapidly discharged into the suppression pool via vertical discharge pipes. This process triggers intense pressure oscillations, which threaten the structural integrity of the containment system, pipe connections, and safety valves. To clarify the underlying physical mechanism and improve prediction accuracy for practical engineering design, a calculation model for steam-air bubble condensation pressure oscillation was established based on bubble dynamics equations. This model specifically incorporates the effects of non-condensable gases on interface heat transfer coefficient and mass transfer resistance, as well as the coupling interaction between bubbles. These factors often overlooked in existing models but critical in nuclear engineering applications. Experimental validation was conducted using a visualized water tank with a discharge nozzle. The nozzle has an inner diameter of 12 mm and an immersion depth of 900 mm. Tests were performed under typical nuclear power plant operating conditions: steam mass fluxes ranging from 70 to 120 kg/(m²·s), liquid subcooling degrees of 40-70 K, and air mass fractions of 10%. High-speed cameras and pressure sensors were used to record bubble morphology and pressure variation data. The results demonstrate that the proposed model reliably predicts bubble formation, growth, coalescence behaviors and pressure oscillation characteristics. The maximum relative errors for bubble equivalent diameter, pressure oscillation frequency, and amplitude are 20%, 13.9%, and 1.9%, respectively. Its prediction performance outperforms existing models that ignore non-condensable gas effects. Further parametric analysis reveals that pressure oscillations are synergistically influenced by steam mass flux and steam-air composition, while liquid subcooling has a negligible effect on oscillation frequency. At a low mass flux of 70 kg/(m²·s), increasing the air mass fraction reduces the dominant oscillation frequency and enhances pressure intensity by slowing bubble condensation and increasing bubble residence time. At a high mass flux of 110 kg/(m²·s), the dominant frequency and pressure intensity first increase with air mass fraction due to enhanced bubble coalescence, then decrease when the air mass fraction exceeds 35% as condensation resistance becomes the dominant factor. This study provides a theoretical tool for the safe design, operation optimization, and risk assessment of suppression pools and containment pressure relief systems in nuclear power plants, supporting the improvement of nuclear reactor safety under accident conditions and reducing the risk of equipment fatigue damage.