Development of Code for Steady-State Thermal-Hydraulic Analysis in Bimodal Space Nuclear Reactor with Heat Pipe

In order to study the core steady-state safety characteristics of Bimodal Space Nuclear Reactor with Heat Pipe（ HP-BSNR）, the thermal-hydraulic models of modified core are established and a steady-state thermal-hydraulic analysis code for the HP-BSNR named STHA_HPBSNR is developed based on the preliminary conceptual design of HP-BSNR. The hydrogen property and steady-state thermal-hydraulic parameters calculated by STHA_HPBSNR are compared with that calculated by program ELM as well as the experimental data in the published literatures, which turn out to agree well. Besides, the effect of different heat transfer and friction resistance correlations on the channel wall temperature is studied. It demonstrates that the STHA_HPBSNR can provide the initial steady-state parameters for the transient-state safety analysis of HP-BSNR.