Numerical Simulation And Analysis Of Sodium-cooled Fast Reactor Fuel Assembly

Nuclear energy has obvious advantages over traditional energy sources.The sodium-cooled fast reactor is a fast breeder reactor that uses liquid sodium as the coolant.As the mainstream reactor type in the fourth-generation fast reactor,the sodium-cooled fast reactor can not only solve the problem The longevity of nuclear waste can also solve the problem of nuclear fuel shortage.The safe operation of nuclear reactors is very important.In order to ensure that the sodium-cooled fast reactor can safely discharge heat in the event of an emergency shutdown,a passive waste heat removal system is generally set up.In the case of a total loss of power,the sodium-cooled fast reactor must rely on the primary circuit and the accident waste heat emission system for cooling after an emergency shutdown.Among them,the natural circulation of the primary loop established through the core is the only way to cool the core fuel.In order to ensure the safe operation and reliable operation of the sodium-cooled fast reactor,it is very important to analyze the natural circulation flow characteristics during the waste heat removal period of the accident.In this study,the natural convection circulation experimental device of core multi-box components was selected as an example,37 rod components were selected as the research object,and numerical calculation software was used for calculation.By comparing and analyzing the calculation results of the friction resistance coefficient of the laminar flow model,the SST k-ω model,the standard k-ε low Reynolds number model and the k-ε double-layer model,And compared the results with the published empirical relationship of frictional resistance coefficients,and selected a suitable flow model to analyze the flow and heat transfer of fast reactor fuel assemblies under natural circulation conditions.So as to better understand its natural circulation mechanism,and provide a powerful reference for the design and optimization of the sodium-cooled fast reactor.Numerical simulation results show that the SST k-ω model has good adaptability,and the model can be used to calculate the natural circulation of multiple cartridge fuel assemblies in the core.Through isothermal flow simulation studies,it is found that the inlet section of the sodium-cooled fast reactor fuel assembly has a greater impact on the frictional resistance coefficient,while the outlet section has a smaller impact;and the length of the inlet section is small and the distance between a pitch,that is,the length of the inlet section is small and 150 mm.By comparing and analyzing the velocity distribution cloud diagrams obtained under isothermal flow and heating conditions,it is found that the velocity distribution cloud diagram under heating conditions is more uniform,and the difference between the fluid velocity in the middle sub-channel and the fluid velocity in the side channels is small.By observing the temperature distribution cloud map of the fuel assembly,it is found that the temperature cloud map shows a trend of high in the middle and low in the periphery.For the side channel and the corner channel area,the flow area is larger and the flow velocity is faster,so The temperature is low.The presence of the winding wire in the rod bundle fully muddled the coolant and made the temperature distribution more uniform.