Joint Cross-Scale Simulation Study Of The Main Cooling System Of A Sodium-Cooled Fast Reactor

As an important component of the sodium-cooled fast reactor system,modelling and simulation of the primary cooling system is useful for rapid simulation and prediction of the flow and heat transfer phenomena in the primary cooling system.The core is the core component of the main cooling system and the study of its thermal and hydraulic characteristics is particularly important.Current programs for fast reactor thermal and hydraulic simulation consist of large scale system and sub-channel programs,and small scale CFD programs.Large-scale programs have advantages in terms of fast system modelling and computational speed,but low resolution and little feedback,while small-scale programs have significant advantages in terms of accuracy,but low computational efficiency.Therefore,in order to balance the efficiency and accuracy of the simulation calculation,it is necessary to carry out joint cross-scale simulation according to the simulation calculation needs of key parts.In this paper,the main components of the fast reactor primary cooling system are modelled in a "modular and parametric" way,and the fast cooling system is modelled and simulated in a fast way.The cross-scale joint simulation and analysis of the reactor main cooling system was carried out.Firstly,this paper establishes a mathematical and physical model of the main cooling system of a pool-type sodium-cooled fast reactor based on the idea of modular modelling,and develops optimisation functional modules such as parametric fast modelling,control body adaptation division and core parallel channel modelling,and completes fast modelling simulations with the pool-type fast reactor CEFR and CFR600 as research objects,verifying the applicability of the design parameters of each component,control body parameterisation and core parallel channel calculation under steady-state operating conditions and the accuracy of the simulation of key parameters of the core module under transient operating conditions.Secondly,the paper addresses the needs of CFD refinement analysis of fast reactor cores,and investigates the modelling of the geometrical model of the wire-wound rod bundle assembly,the meshing process,the turbulence model,the applicability of the turbulent Prandtl number model,the computational control and the post-processing analysis.A parametric modelling tool,automated control calculations,and post-processing of the sub-channel basin of the wire-wound rod bundle have been developed to assist in the refinement of CFD calculations for fast reactor components,and an automatic distributed parallel CFD calculation support platform for fast reactor cores has been designed based on a distributed parallel CFD calculation scheme.Finally,based on the characteristics of the system program and the core CFD program,this paper adopts the overlapping area method and designs a cross-scale joint simulation scheme.The joint simulation control program is written using dynamic link library and FLUENT secondary development techniques,and the joint simulation study of the two types of programs is carried out.While optimising the calculation accuracy of the core parallel channel module of the system program,the refined flow heat transfer state and characteristics of the core fuel assembly rod bundle basin are further investigated by CFD techniques.The above three aspects of research provide supporting technical solutions for the modelling and simulation of the main cooling system of sodium-cooled fast reactors,CFD computational analysis of the core and joint cross-scale simulation,which can be used for reference in the engineering design and safety analysis of fast reactors.