Heat pipe cooled reactors are of interest due to their simplicity of design, scalability, safety and reliability, and are considered ideal for nuclear power generation in space and deep ocean applications. The combined application of supercritical CO2 Brayton cycle system and heat pipes can meet the needs of miniaturization and high energy conversion efficiency of nuclear energy systems. Although scholars have proposed various conceptual design schemes based on this idea, there is a lack of research in evaluating the operational characteristics of these reactors, especially the impact of loss of external load accidents on the operation of the system. Therefore, in this study, a coupled code for a heat pipe cooled reactor and Brayton cycle system is developed for assessing the transient characterization of the heat pipe cooled reactor nuclear system. The reactor part of the code includes a neutron physics model, a heat pipe model, and a core heat transfer model, and it is validated using reference data and experimental results. The supercritical CO2 Brayton cycle system part was modeled using a secondary developed Relap5 program and the coupling of the two parts of the code was completed. The loss of external load accident of the system was simulated using this coupled code and the results showed that the decay heat discharge system and the pre-pressurised containment setup can ensure the safety of the system under accident condition. The coupled code provides can effective tool for the design and safety analysis of supercritical CO2 heat pipe cooled reactors.