Research On Core Flow Distribution Of Natural Circulation Sodium-cooled Fast Reactor

In the case of complete loss of power,the sodium-cooled fast reactor relies on the primary circuit and the accident waste heat removal system for cooling after an emergency shutdown.Among them,the primary loop natural circulation established through the core is the only way to cool the core fuel.The coolant flow rate of the reactor main circuit is mainly determined by the driving force of the circuit.For natural circulation conditions,since the driving force is mainly provided by the density difference between the hot and cold fluids,the total loop circulation flow provided is limited,so the flow distribution at the core inlet directly affects the internal circulation system of the reactor The thermal-hydraulic behavior of the reactor and its good matching with the core power are also conducive to the further improvement of reactor safety performance and economic performance.The research object of this paper is a multi-box component experimental system of a sodium-cooled fast reactor.In order to study the flow distribution of the core,it is necessary to simulate the thermal-hydraulic behavior of the core components on the basis of this experimental system.Due to the complex structure of multi-box components,the use of CFD(Computational fluid dynamics)simulation method requires higher computing hardware conditions.In order to improve the simulation efficiency,this article simplifies the experimental object to a certain extent,and uses the method of programming calculation to perform thermal engineering on the experimental object.Mechanics simulation,comparing the simulation results with experimental data,testing the applicability of the program.The work of the thesis mainly includes:(1)Obtain the corresponding physical model by simplifying the experimental system and establish a mathematical model to solve the system.Discrete the control equations,and develop a thermal-hydraulic solution program for the experimental system through programming.Regarding the simulated core components as closed parallel channels,a one-dimensional flow and heat transfer model is used to develop a parallel channel flow distribution solving program,and the accuracy and reliability of the program calculation are verified by simulating and comparing the operating conditions of the experimental platform.(2)Through the above-developed program,the experimental system is set to calculate different simulated working conditions,and the factors that affect the core inlet flow distribution under the same working condition and different working conditions are explored.The mechanism of flow distribution.(3)Reasonable core flow distribution must not only meet the design safety requirements of the reactor,but also consider improving the performance of the core.Therefore,the optimization of the core flow distribution is conducive to the further improvement of the thermal performance of the reactor.Based on the above research,this paper designs a set of core flow distribution optimization solution method and corresponding calculation program coupled with the above parallel channel flow distribution solution program.Through reasonable adjustment of the inlet resistance,the flow share of each component of the core is matched with the share of power distribution,and the flow distribution optimization program is verified by a simulation example.The task of this paper is to develop an analysis tool for the core thermal safety analysis and thermal hydraulic design of the sodium-cooled fast reactor under natural circulation,and to study the core flow rate of the sodium-cooled fast reactor under natural circulation and small flow conditions.Distribution characteristics.The developed parallel channel flow distribution and flow optimization program has certain academic and engineering significance in order to further enrich the thermal hydraulics research of the sodium-cooled fast reactor and improve the safety and economy of the sodium-cooled fast reactor.