Compared to other energy sources,nuclear reactors offer several advantages as a spacecraft power source,including compact size,high power density,and long operating life.These qualities make nuclear power an ideal ene...Compared to other energy sources,nuclear reactors offer several advantages as a spacecraft power source,including compact size,high power density,and long operating life.These qualities make nuclear power an ideal energy source for future deep space exploration.A whole system model of the space nuclear reactor consisting of the reactor neutron kinetics,reactivity control,reactor heat transfer,heat exchanger,and thermoelectric converter was developed.In addition,an electrical power control system was designed based on the developed dynamic model.The GRS method was used to quantitatively calculate the uncertainty of coupling parameters of the neutronics,thermal-hydraulics,and control system for the space reactor.The Spearman correlation coefficient was applied in the sensitivity analysis of system input parameters to output parameters.The calculation results showed that the uncertainty of the output parameters caused by coupling parameters had the most considerable variation,with a relative standard deviation<2.01%.Effective delayed neutron fraction was most sensitive to electrical power.To obtain optimal control performance,the non-dominated sorting genetic algorithm method was employed to optimize the controller parameters based on the uncertainty quantification calculation.Two typical transient simulations were conducted to test the adaptive ability of the optimized controller in the uncertainty dynamic system,including 100%full power(FP)to 90%FP step load reduction transient and 5%FP/min linear variable load transient.The results showed that,considering the influence of system uncertainty,the optimized controller could improve the response speed and load following accuracy of electrical power control,in which the effectiveness and superiority have been verified.展开更多
With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5(RELAP5), w...With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5(RELAP5), with the implementation of sodium–potassium eutectic alloy(NaK-78) properties and heat transfer correlations, is adopted to analyze the thermal–hydraulic characteristics of the space nuclear reactor TOPAZ-Ⅱ.A RELAP5 model including thermionic fuel elements(TFEs), reactor core, radiator, coolant loop, and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector,moderator, and reactivity insertion effects of the control drums and safety drums are considered. To benchmark the integrated TOPAZ-Ⅱ system model, an electrical ground test of the fully integrated TOPAZ-Ⅱ system, the V-71 unit,is simulated and analyzed. The calculated coolant temperature and system pressure are in acceptable agreement with the experimental data for the maximum relative errors of 8 and 10%, respectively. The detailed thermal–hydraulic characteristics of TOPAZ-Ⅱ are then simulated and analyzed at the steady state. The calculation results agree well with the design values. The current work provides a solid foundation for space reactor design and transient analysis in the future.展开更多
空间熔盐堆运行过程中,反应性控制起到十分重要的作用,目前陆地上常采用的主动反应性控制方式存在出现故障的概率等问题。空间熔盐堆需要一种非能动反应性控制方式,在无需外源及人为操控情况下,对反应性实施自动控制,降低事故发生风险...空间熔盐堆运行过程中,反应性控制起到十分重要的作用,目前陆地上常采用的主动反应性控制方式存在出现故障的概率等问题。空间熔盐堆需要一种非能动反应性控制方式,在无需外源及人为操控情况下,对反应性实施自动控制,降低事故发生风险。本文针对热管式液态燃料空间熔盐堆,利用液态燃料的热胀冷缩机制,当堆芯处在正常运行工况下温度发生变化时,提出并设计一种液态燃料移出移入非能动反应性控制系统(Liquid Fuel in/out Transfer in a Passive Reactivity Control System,LFT-PRCS),并对含有该系统的堆芯进行在正常运行工况下物理特性分析,以及该系统结构参数与反应性补偿能力分析。结果表明:正常运行工况下,含有LFT-PRCS的堆芯具有更负的反应性,且堆芯物理特性未发生明显变化;LFT-PRCS中毛细管道较佳结构参数为:高度为10 cm、内层半径为0.2 cm、外层半径为0.4 cm;LFT-PRCS在寿期初、寿期末温度发生2 K波动时,可向堆芯引入约20 pcm的反应性。上述结果表明,LFT-PRCS可提高堆芯固有安全性,一定程度上补偿燃耗造成的反应性损失。展开更多
基金supported by the National Natural Science Foundation of China(12305185)Natural Science Foundation of Hunan Province,China(No.2023JJ50122)+1 种基金International Cooperative Research Project of the Ministry of Education,China(No.HZKY20220355)Scientific Research Foundation of the Education Department of Hunan Province,China(No.22A0307).
文摘Compared to other energy sources,nuclear reactors offer several advantages as a spacecraft power source,including compact size,high power density,and long operating life.These qualities make nuclear power an ideal energy source for future deep space exploration.A whole system model of the space nuclear reactor consisting of the reactor neutron kinetics,reactivity control,reactor heat transfer,heat exchanger,and thermoelectric converter was developed.In addition,an electrical power control system was designed based on the developed dynamic model.The GRS method was used to quantitatively calculate the uncertainty of coupling parameters of the neutronics,thermal-hydraulics,and control system for the space reactor.The Spearman correlation coefficient was applied in the sensitivity analysis of system input parameters to output parameters.The calculation results showed that the uncertainty of the output parameters caused by coupling parameters had the most considerable variation,with a relative standard deviation<2.01%.Effective delayed neutron fraction was most sensitive to electrical power.To obtain optimal control performance,the non-dominated sorting genetic algorithm method was employed to optimize the controller parameters based on the uncertainty quantification calculation.Two typical transient simulations were conducted to test the adaptive ability of the optimized controller in the uncertainty dynamic system,including 100%full power(FP)to 90%FP step load reduction transient and 5%FP/min linear variable load transient.The results showed that,considering the influence of system uncertainty,the optimized controller could improve the response speed and load following accuracy of electrical power control,in which the effectiveness and superiority have been verified.
基金supported by the China National Postdoctoral Program for Innovative Talents(No.BX201600124)China Postdoctoral Science Foundation(No.2016M600796)the National Natural Science Foundation of China(No.11605131)
文摘With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5(RELAP5), with the implementation of sodium–potassium eutectic alloy(NaK-78) properties and heat transfer correlations, is adopted to analyze the thermal–hydraulic characteristics of the space nuclear reactor TOPAZ-Ⅱ.A RELAP5 model including thermionic fuel elements(TFEs), reactor core, radiator, coolant loop, and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector,moderator, and reactivity insertion effects of the control drums and safety drums are considered. To benchmark the integrated TOPAZ-Ⅱ system model, an electrical ground test of the fully integrated TOPAZ-Ⅱ system, the V-71 unit,is simulated and analyzed. The calculated coolant temperature and system pressure are in acceptable agreement with the experimental data for the maximum relative errors of 8 and 10%, respectively. The detailed thermal–hydraulic characteristics of TOPAZ-Ⅱ are then simulated and analyzed at the steady state. The calculation results agree well with the design values. The current work provides a solid foundation for space reactor design and transient analysis in the future.
文摘空间熔盐堆运行过程中,反应性控制起到十分重要的作用,目前陆地上常采用的主动反应性控制方式存在出现故障的概率等问题。空间熔盐堆需要一种非能动反应性控制方式,在无需外源及人为操控情况下,对反应性实施自动控制,降低事故发生风险。本文针对热管式液态燃料空间熔盐堆,利用液态燃料的热胀冷缩机制,当堆芯处在正常运行工况下温度发生变化时,提出并设计一种液态燃料移出移入非能动反应性控制系统(Liquid Fuel in/out Transfer in a Passive Reactivity Control System,LFT-PRCS),并对含有该系统的堆芯进行在正常运行工况下物理特性分析,以及该系统结构参数与反应性补偿能力分析。结果表明:正常运行工况下,含有LFT-PRCS的堆芯具有更负的反应性,且堆芯物理特性未发生明显变化;LFT-PRCS中毛细管道较佳结构参数为:高度为10 cm、内层半径为0.2 cm、外层半径为0.4 cm;LFT-PRCS在寿期初、寿期末温度发生2 K波动时,可向堆芯引入约20 pcm的反应性。上述结果表明,LFT-PRCS可提高堆芯固有安全性,一定程度上补偿燃耗造成的反应性损失。
