Core thermal-hydraulic evaluation of a heat pipe cooled nuclear reactor

Heat pipe cooled reactor has been proposed for applying in the space station and underwater exploration featured with high reliability, low noise level and compact structure. A typical design of a heat pipe cooled reactor system is HP-STMCs designed by the University of New Mexico. It is conceptually designed as a fast reactor, lithium-filled heat pipes that transfer heat to potassium heat pipes radiator and thermoelectric generators. In this paper, a code is developed for studying the operating thermal–hydraulic characteristics of the reactor core, which consists of core thermal–hydraulic model, neutron kinetics model and heat pipe startup model. In order to verify the code, a series of experiments about potassium heat pipe are operated. Steady-state and startup transient are performed by the code. The highest temperatures of fuel and matrix in reactor core show at the 2nd channel, which reach 1720 K and 1545 K from the results of steady-state. As for the startup transient, the code for heat pipe cooled reactor core successfully predicts key parameters during startup transient, such as each layer material temperature of reactor core, a single heat pipe's temperature response and heat rejection power. The code has initially predicted the transient response of the heat pipe cooled reactor under power increasing rates of 800 W/s and130W/s. It is discussed that the startup strategy of power increasing rates is based on heat pipe transition temperature (916 K). This research provides valuable experience for designing and formulating control strategies of the heat pipe cooled reactor.