The steam generator is one of the most important components of a pressurized water reactor (PWR) nuclear power plant. Thus, the ability to model and predict the steam generator steady-state and transient thermal-hydraulic behavior is a prerequisite for performing safety analyses of PWR systems. A RELAP5 model of the Westinghouse D4 steam generator with a 70/30 split feedwater system has been developed, and it is tested by simulating five secondary-side-initiated transients. This study of primary-to-secondary heat transfer and the secondary coolant vaporization process has enabled the primary coolant cooldown to be maximized, as required for performing a conservative steamline break analysis. These tests were realized using the RELAP5/MOD2.36.05 and RELAP5/MOD3.5M5 computer codes.