Sub-channel analysis for Pb-Bi-cooled direct contact boiling water fast reactor

Abstract

Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR) featured with a direct-contact heat exchanger between lead-bismuth eutectic coolant and water could significantly simplify the primary system and enhance the natural circulation capability, meeting the potential needs for small modular reactor design. It is of great importance to conduct thermal-hydraulic analysis of the PBWFR core in detail. In this paper, a self-developed SUB-channel AnalysiS code SUBAS is adopted to study the thermal hydraulic characteristics of the PBWFR core. The fidelity and the reliability of the code have been preliminarily benchmarked. With SUBAS, the space grid is studied to figure out its impact on the temperature and flow distributions in each sub-channel. Besides, the application of space grids would increase the pressure drops and decreases the cross flow between adjacent sub-channels. To study the transient performance of the PBWFR core, the power transient and the inlet blockage accident are calculated by SUBAS. The results of the power transient show the cross-flow effect would be weakened in the sub-channel which has higher coolant temperature and larger mass flow rate. For the inlet blockage accident, the results indicate the influence of the small area blockage is relatively weak on the overall performance of the assembly but is significant on the local parameters. With consideration of time and space, the blockage influence only exists in a certain area. This research may provide contribution to the design of PBWFR.