Thermal-hydraulic Analysis Code Development for Natural Circulation of Lead-bismuth

Natural circulation of coolant in the leadbismuth reactor is essentially important to remove heat from reactor core under normal or accident conditions, and the thermalhydraulic analysis of natural circulation is a significant work for the designing and licensing. Recently, some code development and validation work for natural circulation analysis of leadbismuth had been carried out around the world. Most of these codes are first developed based on conventional fluids, in which some models are not adequate or necessary for leadbismuth. Owing to the unique features of this coolant and lack of operation experiences, analysis code for natural circulation analysis is still under developing, and a lot of verification work still needs to be done. A thermalhydraulic analysis code suitable for the simulation of transient process of natural circulation of leadbismuth is developed based on the thermalhydraulic physical models of main components in the primary loop of leadbismuth reactors, such as core, heat exchanger, pipe and so on. The single-phase flow model was considered in the code and the momentum equation of the control volume is integrated along the closed natural circulation path to obtain the system momentum equation. Semiimplicit method was used to discretize the momentum and energy equations. In natural circulation analysis, system momentum equation was coupled with energy equation, and these nonlinear and coupled equations are solved iteratively. The temperature distribution of previous time step was used to solve the system momentum equation and the mass flowrate in the flow path was obtained which could be used to solve the energy equation of reactor components along the closed natural circulation path from core inlet. Repeat this process until error of temperature and flowrate between last iteration was less than 10-5. This code was programmed with FORTRAN 95 and adoptd modular structure which includes physical properties module, inputoutput module, reactor component module, thermalhydraulic correlation module and solution module. Preliminary validation was evidenced by using results of startup of natural circulation experiment and power step influence experiment carried out on the natural circulation loop of leadbismuth. According to the result, there is a good agreement between the code predicting and experiment, and the transient process of natural circulation can be well simulated. Relative errors of temperature between calculation results and experimental data are within 5% during the whole transient process, and the relative errors of flowrate are almost 10%. This code can support the analysis work of natural circulation during the development of leadbismuth reactors.