Development of modular dynamic code for steam power conversion system in sodium-cooled fast reactor

Sodium-cooled fast reactor (SFR) is one of the most promising Generation IV reactor types. To improve operational safety and economy, SFR requires accurate prediction of the steam power conversion system. Therefore, a one-dimensional thermal-hydraulic analysis code of the steam power conversion system of SFR is developed in this study where each equipment is packaged as a separate module with modular modeling method. JFNK is adopted to solve ODE in this code, which expand the universality of the code. The model is validated by comparing with design results from the Daya Bay nuclear power plant and the BN-600 steam generator. The maximum relative errors for pressure and temperature are 5.44% and 2.40%, respectively. After validation of the code, the steady-state analysis of the SFR steam power conversion circuit is carried out where the flow and heat transfer process of key equipment can be accurately simulated considering a complete steam power conversion cycle. This work provides reference value for the thermal-hydraulic characteristics of steam power conversion system in SFR.