Around the world, countries are increasingly considering carbon-free energy generation options as the threat of climate change grows. Small modular reactor designs, promising such carbon-free energy generation, are thriving worldwide with novel and innovative technologies that improve safety as well as economic performance. Canada is also considering small modular reactors (SMRs) as a means of achieving net zero carbon emissions by 2050. Some of these reactor designs utilize pressurized water for cooling and moderator. Reactors with pressurized water have been subjected to steam generator tube ruptures in the past, and a detailed investigation into the possible consequences of such incidents in SMRs should be conducted. In this research, a model for one of the newer designs, the NuScale Integrated Small Modular Reactor, was developed with the RELAP5-3D code for assessing safety related transients. The NuScale Small Modular Reactor incorporates helical coil steam generators within its reactor pressure vessel, which are more efficient in terms of heat transfer than the U-tube steam generators that are widely used in nuclear reactors. In the first part of the research, a detailed model is created and used to obtain steady state conditions with parameters collected from NuScale’s Final Safety Analysis Report (FSAR). The Steam Generator Tube Rupture event is analyzed in the second part of the work. Slight differences in the broken and intact steam generator pressures as well as decay heat removal system flow rates are seen in the comparison of reference values and calculated results. Among the reasons for those differences could be that the correlations used by the RELAP5-3D code for heat transfer coefficient and pressure drop in the helical coil steam generators are different than those of the NuScale proprietary code NRELAP5, with which the analyses have been performed in the FSAR. Also, post-dryout heat transfer at the exit of helical coil steam generators and evaporator sections could cause differences in the outlet conditions of the steam, resulting in different mass flow rates as well. The final section of the research simulates a comparable but more severe tube rupture incident without the availability of decay heat removal systems in order to assess the reactor’s emergency core cooling system reaction. Passive decay heat removal systems are crucial components for removing heat after reactor shutdown through heat exchangers that are submerged in the reactor pool and connected to steam generators by a closed loop. The containment pressures, the containment wall temperatures, and the peak fuel clad temperatures are considered to be the key design constraints that must be observed. Future work on this subject could include modifying the source code, adding specific correlations for helical coil steam generators, and comparing the results, as well as quantifying uncertainties in the SGTR event. Main parameters in the quantification of uncertainties would be reactor power, single phase and two-phase discharge coefficients from the break, trip signals and delays as well as break size and location. / Thesis / Master of Applied Science (MASc)