一回路一台泵停运-单环路余热排出工况下CEFR钠池三维瞬态热工特性数值模拟  

Three-dimensional Numerical Simulation on Transient Thermal Characteristics in CEFR Sodium Pool in the Case of One Primary Pump Trip-Residual Heat Removal of Single Loop

作　　者：赵海琦 陆道纲[1,2] 殷晶 梁江涛 杨军 郭忠孝[3] 张钰浩 ZHAO Haiqi;LU Daogang;YIN Jing;LIANG Jiangtao;YANG Jun;GUO Zhongxiao;ZHANG Yuhao(School of Nuclear Science and Engineering,North China Electric Power University,Beijing 102206,China;Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy,Beijing 102206,China;China Institute of Atomic Energy,Beijing 102413,China)

机构地区：[1]华北电力大学核科学与工程学院,北京102206 [2]北京市非能动安全重点实验室,北京102206 [3]中国原子能科学研究院,北京102413

出　　处：《核科学与工程》2022年第6期1277-1284,共8页Nuclear Science and Engineering

摘　　要：一回路一台泵停运-单环路余热排出是池式钠冷快堆的设计基准事故之一,有必要对该工况下钠池内的热工特性进行分析。由于钠池整体尺寸大,难以开展实验研究,通常采用数值模拟的方法进行研究。因此,本研究基于计算流体动力学(CFD)方法,开展了该工况下CEFR钠池三维瞬态数值模拟,得到在一回路泵惰转、返流和非对称余热排出作用下钠池内三维瞬态流动、温度分布以及堆芯出口温度、中间热交换器(IHX)进出口温度等关键参数。计算结果表明,故障环路中泵、IHX存在返流现象。在900 s内,堆芯出口温度降至394.9℃。正常环路IHX出口温度在400 s左右达到最大值360.5℃,随后逐渐降低。故障环路IHX出口温度先下降后上升,900 s时接近364.3℃。具有余热排出的环路具有事故缓解能力,钠池整体温度没有明显升高。研究结果能够为一回路一台泵停运-单环路余热排出事故下池式钠冷快堆安全分析提供参考。One primary pump trip-residual heat removal by single loop is one of the design basis accident of the pool-type sodium-cooled fast reactor,for which the complicated thermal characteristics deserve to be analyzed in depth.It is difficult to carry out experimental study due to the large dimension of the sodium pool,whereas the numerical method is applicable for the simulation.Therefore,three-dimensional transient numerical simulation of China Experimental Fast Reactor(CEFR)sodium pool in the case of one primary pump trip-residual heat removal by single loop accident is carried out based on computational fluid dynamics(CFD)method.The three-dimensional transient flow,temperature distribution,as well as the key parameters including core outlet temperature,intermediate heat exchanger(IHX)inlet and outlet temperature are obtained under the influence of primary pump coast-down,reverse-flow and asymmetric residual heat removal.The calculated results show that the reverse-flow will develop in the pump and IHXs of the fault loop.Meanwhile,the temperature drops to 394.9℃at 900 second in the core outlet.The temperature of the normal loop in the IHX outlet reaches the maximum value of 360.5℃at around 400 second,and then decreases gradually.The temperature of the failure loop in IHX outlet drops firstly and then rises,which reaches about 364.3℃at 900 second.It indicates that the residual heat can be removed by the intact loop to mitigate the accident,so that the overall temperature in the sodium pool can maintain in a relatively low value.It provides important references for the safety analysis of sodium-cooled fast reactor under the one primary pump trip-residual heat removal of single loop accident.

关 键 词：中国实验快堆(CEFR) 一回路一台泵停运 单环路余热排出 三维数值模拟 

分 类 号：TL33[核科学技术—核技术及应用]