Calculation Of Flow And Hfat Transfer Of Steam Generator In Sodium-cooled Fast Reactor

The two most important parts for heat transfer in Sodium cooled fast reactor are the reactor core main vessel in the primary loop and steam generator in the power generation loop.The main vessel plays an important role in containing the entire primary sodium for pool-type sodium-cooled fast reactors(SFRs)while the steam generator is a very crucial component of a power plant system that ensures the safety and efficiency of a plant.The Reactor Vessel Cooling System(RVCS)has a great effect on cooling the main vessel.Simulation analysis for Sodium-cooled fast reactor by ANSYS is proposed to evaluate the thermal-hydraulic characteristics for SFR.ANSYS Fluent is used for analysis that employs the finite element method to achieve the task.The temperature distribution within the reactor core main vessel and steam generator is analysed.In reactor core temperature profile is analyzed along the radius and elevation of core and results of simulation is compared by experimental results of an Experimental study on the main vessel cooling circuit.In this paper I also present a comprehensive overview of CHF enhancement experiments and best approaches to expand thermal margin.This study is directly related to nuclear energy engineering and reactor safety.Describing the different ways like amelioration of fluid properties,effect of flow parameters on CHF,and surface modification which is significant in improving CHF.Also some condition attributes like pipe diameter,circulating mode and flow stability,etc.that affect the CHF at low flow conditions and their diagnostic results that predict the main influencing factors for CHF very well.Flow instability could cause the CHF to reach a minimum and CHF could be improved with the increase of pipe size.Effects of surface orientation and heater material on the heat transfer coefficient and CHF by different materials like Copper(Cu),stainless steel(SS)as well as prototype material(SA508)was also reviewed.The material effect that is indicated by thermal effusivity on CHF suggests that surface with larger thermal effusivity has higher CHF values.Experiments to investigate the CHF position in internally heated vertical annulus flow were also reviewed.It is deducted that the position of CHF in internally heated vertical annulus depends upon the mass flow rate.If the mass flow rate is low then CHF will occur earlier and lower its position.At the end compact results provide theoretical reference for the safety analysis of nuclear plants.