Differences and Analysis of Pressurised Water Reactor Containment Design Using Code ACI 349 and Code ACI 359

The prestressed concrete containment structure constitutes the core protective structure of a nuclear power plant. This paper utilises the prestressed concrete containment vessel (PCCV) of the Hualong Pressurised Reactor 1000 (HPR-1000)—a third-generation pressurised water reactor (PWR)—as the primary research prototype. Utilising ANSYS, a finite element model was established, with key points selected at critical locations such as the dome, cylinder, and base slab for stress analysis calculations. Reinforcement quantification derived from the design methodologies and analytical formulations prescribed in ACI 349 and ACI 359 were compared under various loading conditions. This investigation identified the core discrepancies and influencing factors between the two codes in reinforcement design, alongside a sensitivity analysis to identify key parameters affecting reinforcement design in different structural zones. The results indicate that discrepancies in reinforcement requirements stem primarily from the divergent design philosophies and strength assessment formulations, with this influence outweighing variations in load combinations. Furthermore, significant spatial differences exist in the sensitivity of reinforcement designs for key components to parameters such as the height-to-diameter ratio, shutdown seismic actions, accident pressure, and temperature effects. The conclusions of this study establish theoretical foundations and furnish empirical data to enhance the computational efficiency of prestressed concrete containment design for pressurised water reactor (PWR) facilities, while supporting the alignment of national and international regulatory standards. Furthermore, they serve as a technical reference for advancing nuclear power structural design practices.