Development of Thermal-Hydraulic and Safety Analysis Code for a Heat Pipe Cooled Reactor

Abstract Heat pipe cooled reactors have attracted attention as a promising option for nuclear power generation in space and deep-sea applications due to their simple design, scalability, safety and reliability. Although various models of heat pipe-cooled reactors have been proposed, there remains a gap in rapid and efficient thermal-hydraulic programs for evaluating the operational characteristics and safety of these reactors. In this study, a transient neutron-thermal coupling code was developed to assess the transient characteristics of the core of a heat pipe cooled reactor. This code includes a neutron physics model, a heat pipe model, and a core heat transfer model, and it was verified using reference data and experimental results. Additionally, the code was employed to analyze the operational characteristics of typical accidents, such as reactivity insertion accidents, heat pipe failure, and loss of heat sink incidents. In the reactivity insertion accident, the reactor power stabilized to a new steady state after a period. In the accident of heat pipe failure, the heat from the failed fuel assembly was effectively transferred through adjacent heat pipes. During a loss of heat sink accident, the reactor core safely shut down due to negative feedback, with an initial spike in core temperature followed by a decrease. Throughout all transient processes in these accidents, the fuel pin temperatures maintained a significant safety margin, demonstrating the reactor’s strong safety performance in various accident scenarios. This code thus provides an effective tool for the design and safety analysis of heat pipe cooled reactors.