Nuclear power is an important part of Sweden's energy system and contributes with about 30 % of the supplied electric energy. Interest in new construction is currently high and one type of reactor that may be built is small modular reactors, SMRs. A method that is used to evaluate the safety of a traditional nuclear power plant is the so called probabilistic safety assessment (PSA). An important question is if PSA is applicable to SMRs. This thesis examines differences between a BWRX-300 type SMR and a generic boiling water reactor to investigate if the traditional PSA method is applicable to SMRs and what characteristics that may affect the results.  To investigate the topic, a simplified PSA analysis is carried out at level 1 for five different initiating events: extreme snowfall, loss of cooling accident, loss of main feedwater, loss of offsite power supply and transient. These are then compared to a model of a generic BWR reactor.  The conclusions that emerge from the analysis are that the traditional PSA method can be applied for SMRs of the BWRX-300 type. The features that distinguish the BWRX-300 compared to a generic BWR are mainly that the SMR is smaller in size and power and the influence of the passive safety systems. This results in a lower core damage frequency. Some the areas where the PSA method could be developed for SMRs are the analysis of passive components and passive functions.

The demand for fossil-free energy production is rising due to electrification and increased consumption in the energy system. There are also multiple climate goals to reach, to preserve the possibilities of a sustainable future. A response to this is the BWRX-300, a natural circulation boiling water small modular reactor (SMR) concept developed by GE Hitachi Nuclear Energy. It is currently at the forefront of study for many power utility companies around the world. For decision making it is of interest to investigate the capabilities of new facilities. This Master Thesis work's aim is to study the BWRX-300 reactor's feasibility together with evaluating and optimizing its performance using the core simulation softwares Casmo5 and Simulate5. This is carried out by first verifying Simulate5's natural circulation capabilities by modifying an existing forced-circulation reactor to natural circulation, then comparing simulation results to real world data.  Next a comprehensive model of the BWRX-300 reactor pressure vessel is modelled and validated. Equilibrium cores for 12- and 24-month cycle lengths are then simulated where key reactor performance metrics such as fuel economy, safety margins, axial profiles (of voids and pressure drop) and reactor characteristics are extracted. The effect of different fuel assembly designs in the BWRX-300 reactor core is investigated to find first core design optimums. Furthermore the decay heat removal system in the BWRX-300 is investigated. Lastly the results are used to evaluate the optimal operating mode given the current and future more dynamic projected state of the energy system.  The results show that there are no real technical difficulties while operating the BWRX-300 reactor for 12 or 24 months. The decay heat removal system and core flow characteristics provide abundant coolant flow to maintain long term fuel integrity during both normal and abnormal operation modes. More or less routine core design optimization work is required to obtain sufficient safety margins and improve fuel economy. It is observed that the smaller reactor core requires an increase in average fuel enrichment to maintain criticality throughout the cycle, potentially creating an incentive to raise the current licensing limit. However it is deemed possible to avoid this by conducting further fuel design optimization work.