Transient thermal-hydraulic evaluation of lead-bismuth fast reactor by coupling sub-channel and system analysis codes

Abstract

Abstract The lead-based fast reactor (LFR) is one of the most promising choices of the Generation IV reactors as it has better transmutation and breeding capability, as well as higher safety and economical efficiency. In this paper, the new concept LFR Pb-Bi cooled direct-contact-boiling water fast reactor (PBWFR) is studied. Coupled thermal-hydraulic analysis with the in-house system analysis code SACOL and sub-channel analysis code SUBAS is conducted using the one-direction multi-step coupling method. The unprotected transient of over power (UTOP) and unprotected loss of heat sink (ULOHS) accidents are simulated, and the key parameters are compared with the results by single SACOL, preliminary demonstrating the accuracy and rationality of the coupling method. The transient analysis shows the major concern for PBWFR’s operation is the cladding temperature. In UTOP, the maximum peak cladding temperature (PCT) reaches 834.15 °C, exceeding the cladding temperature limit of 800 °C, which suggests the reactor would be at a risk during UTOP. The maximum PCT is 679.43 °C in ULOHS. Due to the immediate reactor shutdown in ULOHS, the safety margin is relatively larger but the PCT is still higher than the limit in normal condition. Therefore, it is better to further optimize the structure or the operation condition of PBWFR. In addition, more detailed information of the core thermal hydraulic performance can be obtained. In this way, the coupled code could make up the shortage of single channel models and extend the function of the sub-channel analysis code.