| After the uncontrolled lifting of the control rod caused the reactivity insertion accident,when the reactor parameters reached the protection setting value of the safety system,an emergency shutdown was triggered.If the safety system fails and the control rods cannot be inserted into the core,it will result in an Anticipated Transient Without Scram(ATWS)accident.Thermal parameters in the reactor under accident conditions are important to evaluate the transient safety of the reactor,and it is necessary to analyze the overall temperature distribution of the core and the sodium pool.Therefore,studying the three-dimensional distribution characteristics in the sodium pool under the reactivity insertion accident under hypothetical accident conditions is of great significance for the safety assessment of fast reactors.Due to the numerous internal components and complex flow channels,it is difficult to conduct relevant experimental research.Under accident conditions where reactivity is introduced,there may be special three-dimensional phenomena in the sodium pool.At present,domestic and foreign scholars mostly use the traditional one-dimensional system code to study the accident,but only focus on the maximum temperature of fuel pellets and fuel element cladding,without considering the overall three-dimensional transient response in the sodium pool.Therefore,it is necessary to conduct a three-dimensional transient study of the primary system for reactivity insertion accidents.Based on the computational fluid dynamics(CFD)method,this paper took the typical pool-type sodium-cooled fast reactor China Experimental Fast Reactor(CEFR)as the object,and simulated the primary system under the conditions of scram after control rod uncontrolled lifting and control rod uncontrolled lifting combined with no scram.Compared with the simulation results of the verified one-dimensional system code,the change trend of key parameters was basically consistent,and further analyzed the three-dimensional transient characteristics of the sodium pool under accident conditions.After the control rod was lifted out of control,reactivity was accidentally introduced,and the increase in core power caused a rapid increase in core outlet temperature.After triggering an emergency shutdown,the core power and the intermediate heat exchanger(IHX)cooling power decreased rapidly,and the maximum average outlet temperature of the core was 543.4℃.Based on thermal parameters such as the inlet temperature of the hottest fuel assembly,the peak temperature of the fuel element cladding was 650.6℃ through the subchannel code SAC-SUB.Under the condition of no emergency shutdown,the hot sodium flowing out of the core continued to enter the sodium pool,causing the temperature inside the pool to rise slowly.The maximum average outlet temperature of the core was 565.5℃,and the maximum temperature of the fuel element cladding did not exceed 677.7℃.At the same time,the thermal parameters of the key structures inside the reactor under two types of accident conditions were obtained,and the temperature difference between the upper and lower surfaces of the support plate on the cold sodium pool did not exceed the limit value.Compared with one-dimensional system codes,CFD can demonstrate the three-dimensional transient distribution characteristics of the overall and critical structures of sodium-cooled fast reactors and conduct structural integrity evaluations.Comparing the two types of accident conditions,the average temperature at the core outlet,the maximum temperature of the fuel element cladding,and the surface temperature of the critical support structure inside the reactor did not exceed the designed safety limits.The research results provide important three-dimensional numerical references for sodium-cooled fast reactors design and safety analysis. |
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