超临界水冷堆辐照考验回路多尺度热工水力学分析  

Multi-scale thermal-hydraulics analysis on SCWR fuel qualification test loop

作　　者：刘晓晶[1] 熊进标[1] 程旭 LIU XiaoJing;XIONG JinBiao;CHENG Xu(School of Nuclear Science and Technology,Shanghai Jiao Tong University,Shanghai 200240,China;Institute for Nuclear and Energy Technologies,Research Centre Karlsruhe,Karlsruhe 76021,Germany)

机构地区：[1]上海交通大学核科学与工程学院,上海200240 [2]Institute for Nuclear and Energy Technologies,Research Centre Karlsruhe,Karlsruhe 76021,Germany

出　　处：《中国科学：物理学、力学、天文学》2019年第11期43-51,共9页Scientia Sinica Physica,Mechanica & Astronomica

基　　金：上海市科技创新行动计划政府间国际科技合作项目(编号:17580711400)资助

摘　　要：作为第四代核能系统中唯一的水冷反应堆,超临界水冷堆(SCWR)具有系统简单、热效率高、经济和安全性好等优点.中国和欧盟联合发起了第七框架研究计划国际合作项目"超临界水冷堆燃料性能验证实验(Supercritical Water-cooled Reactor Fuel Qualification Test, SCWR-FQT)",该实验将对超临界水环境下的小型燃料组件进行性能测试,为实验回路进行设计、分析和验证,并为其设计分析和安全许可申请提供支持.本文从计算流体力学、子通道、系统安全角度出发,对整个系统进行多尺度的热工水力安全分析.本文还对子通道程序COBRA和系统程序ATHLET中的传热模型、摩擦阻力模型和湍流交混模型等进行了修改,使其适用于超临界水冷堆模拟.另外,本文通过交换堆芯出口压力、冷却剂进口温度、堆芯进口冷却剂流量、活性区的产热和摩擦因子等参数实现程序耦合,以得到更为精细的热工水力行为.结果表明,修改过后的程序适用于超临界水回路瞬态分析,现有的安全系统设计可保证组件实验段在事故情况下得到有效冷却. CFD计算结果表明,绕丝对棒束流体传热有较大影响;子通道结果表明角通道堵塞条件下包壳温度最高.同时,计算也证实了多尺度耦合程序可精准预测事故进程和参数分布.As the only one design using water as coolant in Generation IV Nuclear Energy Systems proposed by Generation IV International Forum(GIF), supercritical water cooled reactor(SCWR) has been widely investigated due to its advantages of simplicity, high thermal efficiency, good economy and safety. The Sino-Euro corporation project Supercritical Water Reactor-Fuel Qualification Test(SCWR-FQT) is proposed to analyze and verify a supercritical water-cooled experiment loop containing a small scale fuel assembly and provide support to this loop’s design and license application. To solve the problems above, this paper will carry out the preliminary safety research for the SCWR-FQT facility. Several thermalhydraulic codes(including CFD code, subchannel code, system code) are selected to perform the safety analysis in the facility. COBRA and ATHLET have been modified by implementing the heat transfer model, turbulent mixing and friction models which are applied to supercritical water cooled reactors into the previous code. To determine the detailed thermal-hydraulic behavior in the bundle region at some transient condition, the multi-scale coupled code has also been developed by transferring the pressure at the core outlet, coolant temperature at the core inlet, coolant mass flow rate at the core inlet, heat generated in the active core and pressure friction coefficient of SCWR-FQT. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test and fuel bundles can be cooled effectively with safety measures in accidental conditions. The CFD results show that wire-wrap has large influence on heat transfer. The subchannel code presents that the cladding temperature will be highest when blockage occurs in the corner subchannel. Meanwhile, the coupled results show that the coupled code can predict the accurate accident procedure and detailed parameters distribution.

关 键 词：超临界水冷堆 实验回路 热工水力 多尺度 计算流体力学 

分 类 号：N[自然科学总论]