石墨慢化通道式熔盐堆的稳态热工水力计算模型

针对石墨慢化通道式熔盐堆的堆芯结构,基于COMSOL Multiphysics程序和MATLAB程序建立了堆芯稳态热工水力学计算模型。该模型对堆芯内固体区域的温度分布采用三维热传导方程进行模拟,对通道内熔盐温度采用一维单相流体模型进行计算。固体区域与熔盐通过熔盐通道壁面的对流换热边界建立热耦合。该模型基于平行通道压力损失相等的原则,分配堆芯内各熔盐通道的流量。通过对比RELAP5程序的计算结果,验证了模型对温度和流量分配计算的正确性。针对2 MWt 液态燃料熔盐堆的一种概念设计,分析了堆芯内三维温度分布和通道间流量分配。该模型可精确计算通道式熔盐堆堆芯内稳态温度分布和流量分配,对堆芯的热工水力学设计具有重要意义。     

中国实验快堆全堆芯流量分配计算与试验

针对中国实验快堆(CEFR)堆芯和一回路的设计特点,开发水力特性计算程序DAEMON,完成不同工况下的全堆芯流量分配计算,给出流量分配不均匀性等参数。在反应堆调试阶段,进行全堆芯流量分配试验。结果表明,程序计算值与试验值符合较好。在此基础上,验证了CEFR堆芯的流体力学设计,并为反应堆调试和运行提供了基础数据。     

自然循环反应堆流量分配优化研究

由于自然循环反应堆一回路产生的驱动力有限,回路循环总流量较小,因此堆芯流量分配设计与优化非常重要。合理的堆芯流量分配不仅能满足热工安全要求,还能直接提高堆芯的性能。基于以上原因,本文对自然循环反应堆流量分配优化问题进行了初步研究,对闭式并联通道,采用一维流动传热模型,建立了入口阻力系数优化初值求解模型并设计了精确解搜索算法,并耦合堆芯热工分析程序COBRA编写了相应的堆芯流量分配优化程序。选择一自然循环反应堆算例,采用该程序对堆芯寿期内的流量分配优化进行了计算和分析。结果表明,将各典型寿期节点流量分配优化得到的入口阻力优化设置方案取平均值,可获得相对整个循环寿期达到较好优化效果的入口阻力设置方案。针对取平均值这种人工设计方法难以获得全局最优解的缺点,参考现代优化计算方法,提出了一种自动实现循环寿期内流量分配最优化的方法。     

自然循环反应堆流量分配优化研究

由于自然循环反应堆一回路产生的驱动力有限,回路循环总流量较小,因此堆芯流量分配设计与优化非常重要。合理的堆芯流量分配不仅能满足热工安全要求,还能直接提高堆芯的性能。基于以上原因,本文对自然循环反应堆流量分配优化问题进行了初步研究,对闭式并联通道,采用一维流动传热模型,建立了入口阻力系数优化初值求解模型并设计了精确解搜索算法,并耦合堆芯热工分析程序COBRA编写了相应的堆芯流量分配优化程序。选择一自然循环反应堆算例,采用该程序对堆芯寿期内的流量分配优化进行了计算和分析。结果表明,将各典型寿期节点流量分配优化得到的入口阻力优化设置方案取平均值,可获得相对整个循环寿期达到较好优化效果的入口阻力设置方案。针对取平均值这种人工设计方法难以获得全局最优解的缺点,参考现代优化计算方法,提出了一种自动实现循环寿期内流量分配最优化的方法。     

基于CFD方法的自然循环反应堆冷却剂流动特性分析

运用计算流体力学(CFD)方法对自然循环反应堆冷却剂流场进行计算模拟。分析发现,采用剪切应力传输模型(SST)与雷诺应力模型(SSG)进行流场计算,其结果基本一致;CFD流量计算结果与系统程序Relap5流量计算结果基本一致。对反应堆堆芯采用闭式通道和设置提升筒的方案进行论证分析发现,两种方案均对流量分配产生影响,使流量分配特性与堆芯径向功率分布更加接近,从而实现流量分区。     

秦山核电站二期反应堆堆芯流量分配数值分析

获得可靠的堆芯入口流量分配数据是改善压水堆堆芯热工水力性能的需要。应用计算流体力学方法研究了反应堆压力壳内复杂的流动现象,得到了秦山核电站二期600 MW反应堆1/4整体水力模型的堆芯入口流量分配状况,并对下腔室几何结构、冷管段入口流量等影响因素进行了敏感性研究。分析结果表明,双环路工况入口流量对堆芯入口流量分配影响较小;与双环路工况相比,单环路运行工况时流动特性显著变化,导致流量分配状况差异较大;堆芯入口流量再分配因子为0.05,与原型设计参数吻合。计算结果证实所采用研究方法有效,可为相关反应堆工程设计验证提供依据。     

CAP1400反应堆堆芯入口流量分配试验研究

反应堆堆芯入口流量分配是反应堆水力性能研究的重要内容之一,其与堆芯热裕量和燃料组件燃料棒的流致振动密切相关,从而影响反应堆的运行。CAP1400反应堆堆芯入口流量分配试验是验证CAP1400反应堆结构设计与分析的一个重要环节,旨在验证CAP1400反应堆堆芯入口流量分配的均匀程度。本文通过1/6比例模型试验,获得无均流板结构工况和带均流板结构3种工况（均匀流量工况、非均匀流量工况、偏回路流量工况）下CAP1400反应堆堆芯入口流量分配结果,并进行了各工况下流量分配均匀程度的分析。试验结果表明,CAP1400反应堆堆芯入口具有较好的流量分配效果。     

SCWR堆芯稳态物理-热工水力耦合计算程序系统CASIR的开发

针对超临界水冷反应堆(SCWR)堆芯冷却剂密度沿轴向变化剧烈的特点,开发用于SCWR堆芯稳态物理-热工水力耦合计算的程序系统CASIR。CASIR由改进的压水堆堆芯中子学计算程序和适用于SCWR燃料组件计算的子通道热工-水力程序组成,具备调整堆芯下腔室入口流量分配的功能。针对CSR1000双流程的SCWR首循环堆芯,通过与蒙特卡罗程序对比寿期初时刻计算结果的方式,初步验证CASIR计算SCWR堆芯中子学问题的准确性;通过SCWR堆芯燃耗模拟,以及调整堆芯流量分布使得最大包壳表面温度(MCST)满足设计限值的测试,表明CASIR满足SCWR堆芯设计的要求,可应用于方形燃料组件的SCWR堆芯概念设计。     

基于CFD的钠冷快堆子通道绕丝搅浑系数选取

钠冷快堆能够提高铀资源的利用率,减少核废料的产生,是非常有前景的第四代核能系统堆型之一。同时,钠冷快堆也因其使用金属绕丝对燃料棒进行固定,具有更复杂的堆内构造,探究钠冷快堆堆芯内因绕丝而引起的搅浑效应对钠冷快堆的堆芯设计及安全分析具有重要意义。本文针对钠冷快堆的堆芯设计,采用CFD软件建立带绕丝的7根燃料棒束模型,针对大流量工况（工况1）、中流量工况（工况2）进行工况计算,根据流场的雷诺应力获得绕丝的湍流搅浑系数。并基于自主研发的子通道计算程序SAC-SUB建立相同的几何模型,将湍流搅浑系数输入子通道计算程序中,获得内通道、边通道、角通道温度分布,并将两种软件的计算结果进行了对比。对比结果表明,对于不同的通道而言,两种计算软件内通道的温度偏差最小（2.5℃）,角通道的温度偏差最大（13.2℃）。对于不同的流量而言,中流量工况（工况2）温度偏差更小,最小温差只有0.8℃。该工作为后续快堆子通道分析搅浑系数的选取提供了技术基础。     

并联通道瞬态流量分配方法研究

应用多通道模型对闭式燃料栅格反应堆进行热工水力分析时,首先需要解决流量分配问题。本文提出了3种流量分配方法,编写了瞬态流量分配程序,求解了1个并联通道流量分配问题,并对这3种方法做了计算对比。结果表明,方法1只适合流量缓慢变化工况;而方法2和方法3适用于流量剧烈变化工况;但在计算同一工况时方法3比方法2更稳定;因此,方法3可作为板状燃料堆芯冷却剂通道流量分配的计算方法。     

中国先进研究堆稳态热工水力计算程序开发

针对中国先进研究堆(CARR)的具体特点开发了堆芯多通道热工水力计算程序ECARR。通过对全堆芯的数值模拟，得到了堆芯流量分配和非对称冷却条件下板状燃料元件的温度场，为进一步分析燃料元件的温差热应力等其它参数提供了所需数据。同时还对堆芯最热通道进行了热工水力计算及相应准则的判定。各参数符合CARR热工水力设计准则要求。     

CARR应急堆芯冷却系统停堆冷却措施分析

停堆后的冷却问题是中国先进研究堆(CARR)重要的安全问题之一。CARR应急堆芯冷却系统是一套多功能、高度安全可靠的专设安全设施,它在反应堆正常运行时执行池水冷却功能;在正常停堆和事故停堆过程中执行应急堆芯冷却功能;还执行应急热阱选择、系统供电方式、回路阻力分析、阀门开关设置等方面的处理,使系统在两种功能的切换中不需要人为操作,依靠流量的自动匹配来满足正常运行和事故运行的要求。体现了CARR的安全性、先进性和经济性。本文以核安全法规和导则为前提,以满足系统功能为基础,首先介绍了CARR应急堆芯冷却系统的功能、主要参数和流程。根据CARR的实际情况,对应急堆芯冷却系统的停堆冷却措施和典型事故进行了分析,论证了该系统是如何在正常停堆和事故停堆状态下实现非能动堆芯冷却的。     

Analysis on natural circulation capacity of the CARR

The investigation on natural circulation （NC） characteristics of the China Advanced Research Reactor（CARR） is very valuable for practical engineering application and also a key subject for the CARR. In this study, a computer code was developed to calculate the NC capacity of the CARR under different pool water temperatures. Effects of the pool water temperature on NC characteristics were analyzed. The results show that with increasing pool water temperature, the NC flow rate increases while the NC capacity decreases. Based on the computation results and theoretical deduction, a correlation was proposed on predicting the relationship between the NC mass flow and the core power under different conditions. The correlation prediction agrees well with the computational result within ±10% for the maximal deviation. This work is instructive for the actual operation of the CARR.     

Development of a thermal-hydraulic analysis code for CARR

The China advanced research reactor (CARR) being built in Beijing, China, is a multipurpose research reactor for a variety of fields. Theoretical calculation of thermal hydraulic characteristics of CARR is presented in this paper. The theoretical analysis consists of initial steady and transient accidental analyses. Point reactor neutron kinetics model with six groups of delayed neutron is adopted for the solution of reactor power. All possible flow and heat transfer conditions are considered and the corresponding optional models are supplied in the theoretical calculations. A new simple and convenient model is proposed for the resolution of the transient behaviors of main pump instead of the complicated four-quadrant model. Gear method and Adams predictor–corrector method are adopted alternately for a better solution to such ill-conditioned differential equations corresponding to detail process. The initial multi-channel analysis shows that the effects of geometrical size on flow distribution play dominant role and the effects of core power distribution may be neglected. The temperature fields of fuel elements under asymmetrical cooling condition are also obtained, which are the bases for further study on transient-induced stress analysis, etc. Accidental analyses show that the activity of emergency cooling system apparently reduces the peak temperatures of fuel and coolant, peak quality and other operation parameters. Thus it effectively ensures the safety in operation of CARR. Because of the adoption of modular programming techniques, this code is expected to be applied to accidental analysis of other types of reactors by easily modifying the corresponding function modules. Also, this code is expected to be validated against experimental data.     

压水堆下腔室流场对堆芯流场和温度场影响数值分析

利用三维数值模拟,对不同环腔厚度和环腔内冷却剂速度条件下,下腔室内冷却剂流场进行了计算。在此基础上,对压水堆流量孔板冷却剂流量的分配情况进行了分析,并找出了通过流量孔板通孔小组的冷却剂流量与平均流量的最小偏差。分别计算了最小偏差条件下与平均流量条件下,堆芯内板状燃料元件周围冷却剂的流场和温度场。发现由环腔厚度或环腔内冷却剂速度不同而引起的下腔室流场分布不均匀,对堆芯内冷却剂流场和温度场影响较大。     

中国先进研究堆矩形通道流场数值计算分析

通过SIMPLE数值方法,编制程序,对中国先进研究堆(CARR)全流道进行流场数值模拟.采用对CARR的单个冷却剂通道进行单相水的数值传热计算,并递增地改变流道入口流速,计算获得与入口流速对应的流道速度场与温度场分布,展现其变化规律,分析入口流速对流道热工水力参数分布的影响.采用所编制的程序,对板式燃料组件构成的窄矩形通道进行数值模拟,由此来确定热工水力设计需要的一些反应堆安全参数.这些安全参数为反应堆事故监测系统提供必要的热工过程状态信息,也为CARR提供必要的数据参考.     

压水堆下腔室流量分布数值分析

建立了压水堆下腔室流场的三维数值计算模型,计算了不同环腔厚度和环腔内冷却剂速度条件下,下腔室内冷却剂的流场,分析了环腔厚度和环腔内冷却剂速度对下腔室流向堆芯的流量分布的影响。入口速度不同或环腔厚度不同,在下腔内冷却剂流动形成漩涡的位置、大小和流动速度均会发生改变,导致通过流量孔板通孔的流量分布不同。入口速度较低时,流量孔板上所有通孔的流量分布比较均匀,在平均值附近波动,流量最高的通孔小组出现在边缘处;入口速度较高时,流量明显地呈现出中心高边缘低的特点。通孔小组的流量最大值随着环腔厚度增加由孔板的中心向边缘移动。     

Numerical simulation and experimental research on CHF characteristics of AP1000 nuclear reactor

The present paper aims to investigate the critical heat flux (CHF) characteristics of AP1000 reactor based on the experimental and numerical researches, under normal operation and loop fault conditions, respectively. The differences of flow characteristics in these conditions were analyzed. It indicated that the flow features are very complicated in three dimensions and AP1000 has better self-regulation capability to distribute coolant flow compared to conventional reactors. Under normal operation condition, coolant of two loops is distributed along circumference of the reactor annular channel symmetrically. In case that one of the loops fails suddenly and the coolant is partially lost to total loss, the core flow distribution plate and lower grid plate cannot eliminate uneven flow immediately due to loop failure, also the nonuniformity of reactor coolant flow distribution increases gradually, which leads to the heat transfer deterioration easily. In addition, the reactor core departure from nuclear boiling ratio (DNBR) and CHF does not show a certain linear relation, and the DNBR and CHF of AP1000 are greater than that of conventional reactors which not only improve the reactor thermal efficiency, but also obviously reduce the probability of CHF phenomenon appear.     

Two-phase instability analysis in natural circulation loops of China advanced research reactor

In this paper, two-phase flow instability in natural circulation loops of China Advanced Research Reactor (CARR) has been investigated. CARR is a low pressure and low power density research reactor. A natural circulation instability analysis model is developed for the natural circulation loop of CARR. The homogeneous flow model is used to establish the system control equations. The non-uniform heating and subcooled boiling heat transfer is included. The accumulation heat of the wall is also included. Numerical method of Gear is employed to solve the system equations documented in terms of ordinary differential equations. According to the calculation results, stability maps of the natural circulation loop, which confirm the presence of an instability region under the conditions of low equilibrium quality in the outlet and low pressure, are obtained. It is a special kind of density wave oscillation (DWO) that occurs in very low equilibrium quality region with the characteristics of geysering and ‘Type-I’ DWO at the same time. The calculation results show such oscillation course clearly. The variations of the mass flow rate, the pressure drop and the boiling boundary are analyzed separately. Especially, the phase-space trajectory of the boiling boundary and the mass flow rate is discussed. Finally the oscillation frequency is discussed. The calculated results have important significance for the safety operation and accidental analysis of CARR.