流道方位及运动对沸腾两相流动传热行为的影响

基于两流体模型与固壁非稳态导热模型,结合有关关联式组合与参数综合比选下的模型验证,建立分析流道内沸腾流动传热的瞬态数值模拟程序。通过引入运动下附加加速度模型,研究流道形位与运动等因素对流道内沸腾流动与传热的影响。诸因素对沸腾传热系数、流动压降、固壁温度以及传热流动的瞬态行为影响的计算分析结果对相关堆芯流道的试验设计与应用具有指导意义。     

垂直上升管内对流沸腾瞬态数值模拟及其两相流不稳定性预测

核电站蒸汽发生器二次侧为两相对流沸腾换热过程,在设计过程中须保证其不发生两相流不稳定性。本工作采用时域法对垂直上升管内两相流不稳定性进行研究,建立了垂直上升直管内流动沸腾过程的一维模型,并编制计算程序。采用该程序模拟了流动沸腾过程气液两相流密度波的不稳定性,给出两相流波动过程瞬态参数分布,由此分析了密度波不稳定发生的机理,并分析了质量流速、系统压力、入口过冷度对不稳定的影响。结果表明,与已有实验及理论结果相比,瞬态参数计算结果与实验结果符合较好,可较好找到不同工况下直管内气液两相流发生不稳定的边界,结果优于Khabenski线算图方法。     

严重事故下反应堆压力容器下封头耦合烧蚀传热分析

核电站严重事故发生后,反应堆压力容器（RPV）固壁在熔池作用下会发生烧蚀、减薄。开展RPV下封头耦合烧蚀传热分析对堆坑注水有效性论证和RPV剩余壁厚确认有重要的理论指导意义。本文以CPR1000反应堆压力容器为研究对象,在FLUENT 17.2平台下,基于动态网格方法和UDF二次开发,构建了综合考虑RPV固壁瞬态烧蚀与导热、RPV内壁热流密度再分布及RPV外壁过冷沸腾的全耦合计算模型,获取了9 000 s内的堆坑两相流场分布和RPV固壁烧蚀温度场,分析确定了最小剩余壁厚和发生位置。结果表明：使用动态网格捕捉壁面烧蚀的方法可行,本文全耦合计算模型在分析RPV固壁瞬态烧蚀过程方面有一定优势。     

基于两相漂移流理论的蒸汽发生器计算程序(CCM)的应用

介绍了一个建立在两相漂移流和相边界可连续移动模型基础上的、可较为全面描述单相及两相流的传热和流动的程序CCM.该程序通过"修正的有限元方法"对控制体节点之间建立联系,保证模拟效果的同时,节省了计算时间.以CCM为基础的UTSG-3程序可较好地模拟U型管蒸汽发生器的稳态和瞬态过程,可得到更为广泛的应用.通过不同运行参数的扰动研究了大亚湾核电站蒸汽发生器动态响应过程,结果显示UTSG-3程序具有模拟U型管蒸汽发生器动态行为的能力.     

两流体模型在环形窄缝内的应用研究

采用EPRI最新开发的Chexal-Harrison相壁相间摩擦模型和简化的相壁相间传热模型,构造了适用于环形窄缝内沸腾传热和流动的两流体模型,并编制了热工水力计算程序——THYME程序.与实验数据比较,分析了环形窄缝套管在不同负荷下Relap5/Mod3.2程序和本文程序的计算结果.计算结果表明,Relap5/Mod3.2低估了环形蒸发管的蒸汽温度,本文计算结果与实验数据较为一致.     

基于OpenFOAM的过冷流动沸腾数值模拟

过冷流动沸腾现象被广泛应用于工业生产和动力系统中,对该现象的准确预测是两相流CFD模拟的重要研究方向。本文详细阐述了该模拟过程中的欧拉两流体模型及相关辅助模型,基于开源CFD平台OpenFOAM,模拟了4.5 MPa下竖直圆管内的过冷流动沸腾,得到了截面空泡份额、液相平均温度及壁面温度沿轴向的分布。计算结果与实验值符合良好,说明了模型的有效性和程序的正确性。本文可为在OpenFOAM中添加新的模型及开发新的求解器以模拟过冷流动沸腾问题提供参考。     

基于OpenFOAM的过冷流动沸腾数值模拟

过冷流动沸腾现象被广泛应用于工业生产和动力系统中,对该现象的准确预测是两相流CFD模拟的重要研究方向。本文详细阐述了该模拟过程中的欧拉两流体模型及相关辅助模型,基于开源CFD平台OpenFOAM,模拟了4.5 MPa下竖直圆管内的过冷流动沸腾,得到了截面空泡份额、液相平均温度及壁面温度沿轴向的分布。计算结果与实验值符合良好,说明了模型的有效性和程序的正确性。本文可为在OpenFOAM中添加新的模型及开发新的求解器以模拟过冷流动沸腾问题提供参考。     

湍流模型对瞬态流固耦合换热数值模拟影响分析

采用ANSYS Fluent程序建立了固体加热功率指数增长的气体冷却模型,分析了瞬态过程中的流动传热特性,通过数值模拟结果与3组实验数据进行对比,研究了标准k-ε模型、标准k-ω模型、过渡SST模型和RSM模型4种典型湍流模型对瞬态过程中流固耦合换热数值模拟结果的影响。通过分析发现：热负载指数提升过程中,热功率一部分用于对流换热,而另一部分仍留在固体内,且热功率提升速率越高,对流换热的占比越低;采用不同湍流模型模拟出的等效表面热流密度均低于实验值,且采用ε类湍流模型的模拟结果与实验数据更接近。     

AP1000全失流事故DNBR计算分析

AP1000反应堆冷却剂强迫流动全部丧失事故(简称全失流事故)可能导致堆芯发生偏离泡核沸腾(DNB)。使用美国核安全管制委员会(NRC)的TRACE程序和法国子通道分析程序FLICAⅢ-F,对AP1000电厂系统和堆芯进行建模,使用TRACE程序给出的全失流事故瞬态参数作为FLICAⅢ-F程序的输入条件,进行全失流事故DNB分析。计算结果表明:在瞬态过程中,堆芯内偏离泡核沸腾比(DNBR)数值始终高于安全分析限值,满足DNB设计基准。通过与安全分析报告中的计算数据进行对比,证明本文用TRACE和FLICAⅢ-F程序建立的DNB分析计算模型是合理的,能够用于AP1000电厂的工程设计。     

基于流固共轭传热的两相流动数值模拟及燃料子通道CHF预测研究

为对过冷沸腾两相流动进行准确模拟,并探索临界热流密度（CHF）预测方法,本文基于共轭传热和两相CFD分析的方法,通过流固界面耦合,建立流固共轭传热两相流动耦合求解的数值模型。首先通过典型燃料棒栅元过冷沸腾两相流动的模拟,验证数值模型的正确性。随后对燃料子通道内两相流动进行模拟,并在两相流动模拟的基础上,通过准瞬态的方法,建立与CHF试验过程非常近似的CHF预测方法,将加热壁面的温度飞升作为CHF判定的标准,实现对燃料组件子通道CHF的数值预测。研究表明,本文建立的数值模拟方法,可为燃料组件或其他换热系统的CHF预测奠定基础,为燃料组件的设计提供新的辅助手段。     

Impact of Pseudocolloid Formation on Migration of Nuclides within Fractures

Numerical simulation of boiling of Na-K eutectic alloy (22% Na; 78% K) in a parallel channel system under natural circulation conditions is presented in this work. The calculations pertain to an experimental analysis conducted at the AR-1 facility of IPPE (Russia), analyzing the reactor core coolabilty under severe thermal conditions. The geometrical and operational characteristics reproduce those typical of Russian FBRs. A thermohydraulic code system based on a subchannel analysis code has been developed and adopted for the simulation. Comparisons show that the adopted numerical procedure is capable of reproducing the boiling phenomena inside the parallel bundle test section, correctly predicting the heat transfer conditions prior to and during the boiling. Experimental as well as computational results show the marked influence of the parallel channel interaction, evidenced by the appearance of severe flow oscillations. The results confirm that stable heat removal is provided throughout the entire transient.     

基于RELAP5和FLUENT的多尺度耦合分析方法研究

基于二次开发得到的铅冷快堆一维系统程序RELAP5_LEAD和三维计算流体力学程序FLUENT,利用动态链接库技术和FLUENT用户自定义函数,开发了多尺度耦合分析程序RELAP5/FLUENT。在单相范围内,分别利用耦合程序RELAP5/FLUENT开展简单铅冷串联管道的瞬态流动和传热模拟、简单铅冷闭式回路的瞬态流动模拟,并与RELAP5_LEAD计算结果开展Code-to-Code对比分析。研究结果表明,RELAP5/FLUENT计算结果与RELAP5_LEAD模拟结果吻合良好,耦合程序的开发取得了初步成功,可用于分析铅冷快堆堆内的复杂三维热工水力现象。     

基于CFD的池式快堆多物理耦合分析方法研究

Based on the critical/subcritical point kinetics model, the fuel pin heat transfer model, and the auxiliary thermal hydraulic models such as the heat exchanger model and the porous media model, a multi-physical coupling code CFD/PF was developed by means of the explicit iteration method, dynamic link library technique (DLL) and user-defined functions (UDF) of FLUENT. The CFD/PF was used to carry out the simulation of SNCLFR-100 unprotected transient of over power (UTOP) of a small natural circulation LBE cooled fast reactor, and the code-to-code comparison analysis was conducted with the renowned multi-physical coupling code SIMMER-III. The results indicated that the CFD/PF simulation results are in a good agreement with SIMMER-III calculation results, and the multi-physical analysis method and code development have been achieved initial success, which can be used to analyze the complex three-dimensional flow and heat transfer phenomena in pool-type fast reactors.     

基于CFD的池式快堆多物理耦合分析方法研究

基于临界/次临界点堆中子动力学模型、燃料棒传热模型、热交换器和多孔介质等辅助热工水力模型,采用显式迭代和动态链接库技术（DLL）,利用商用计算流体力学（CFD）程序FLUENT的用户自定义函数（UDF）实现中子动力学、燃料棒热传导等和快堆堆池冷却剂流动换热的耦合计算,开发池式快堆多物理耦合计算程序CFD/PF。采用CFD/PF开展小型自然循环铅铋快堆SNCLFR-10无保护超功率事故（UTOP）模拟,并与国际知名快堆多物理耦合分析程序SIMMR-III的计算结果开展Code-to-Code对比分析。研究结果表明:CFD/PF与SIMMER-III的分析结果吻合良好,耦合程序的开发取得了初步成功,可用于分析池式快堆堆池内的复杂三维流动和换热现象。      

Coupled Ablation and Heat Transfer Analysis of Lower Head for Reactor Pressure Vessel under Severe Accident Condition

After a reactor core melts accident, the solid wall of the reactor pressure vessel (RPV) will be inevitably eroded by the melting core which contains large density of heat flux. The analysis of the coupled ablation and heat transfer of the lower head for RPV is of great theoretical significance to the effectiveness demonstration of water injection in reactor pit and the confirmation of the residual wall thickness of RPV. In this work, numerical simulations were carried out based on the RPV model of CPR1000 using the CFD software FLUENT 17.2. Based on dynamic mesh model and user-defined function (UDF) redevelopment, a fully coupling calculation model considering the transient ablation and heat conduction of solid wall of RPV, the redistribution of heat flux density in RPV inner wall and the subcooled boiling of RPV outer wall was established. Both two-phase flow pattern in the reactor pit and temperature field of RPV solid wall ablation within 9 000 s were obtained and the minimum residual wall thickness and the occurrence location were determined by analysis. The results show that it is feasible to use dynamic mesh to capture wall ablation. The fully coupling calculation model has certain advantages in analyzing the transient ablation process of RPV under severe accident.     

自然循环流动不稳定条件下的传热特性实验研究

为探究流动不稳定性机理,在低压自然循环系统中开展了一系列相关实验,分析了不同流量振荡模式下自然循环的沸腾传热机制及局部传热特性。实验表明：中、低热流密度下出现的较规则的周期性振荡由加热段内流动沸腾诱发,壁面过热度不会随流量振荡而大幅度变化;高热流密度下自然循环系统出现的周期性不规则振荡现象中,流动沸腾类型间的相互转变不是流量波动的唯一原因。大幅度的流量脉动可能在高热流密度下导致沸腾临界的发生,出口壁面出现间歇性干涸,局部传热系数下降的同时伴随壁温的短暂飞跃。随着热流密度的提高,自然循环系统可能出现持续性干涸。     

基于CFD方法的矩形窄缝通道内过冷流动沸腾模型研究

板状燃料元件中的矩形窄缝通道具有宽高比大的几何特征,高度方向速度梯度大、分布陡峭,发生过冷沸腾时,近壁面汽泡运动行为将受其影响而改变,其中汽泡滑移现象对沸腾换热影响较大。本文针对矩形窄缝通道中的汽泡滑移行为,构建了包含滑移热流的壁面热流分配模型,并建立机理性的汽泡受力模型和滑移模型计算汽泡脱离直径、浮升直径和滑移距离等辅助参数,开发了一套适用于矩形窄缝通道内向上流动沸腾的壁面沸腾模型。选用Nuthel窄缝通道沸腾实验进行数值模拟验证,结果表明：本文模型可以较好地预测1～4 MPa中低压工况窄缝通道向上流动沸腾的壁面过热度,最大误差相比RPI模型由80%降低至17%;蒸发热流份额和近壁面空泡份额相比RPI模型更低。     

Phenomenological investigation of sodium boiling in a SFR core during a postulated ULOF transient with CATHARE 2 system code: a stabilized boiling case

In the framework of the research and development on GEN IV sodium fast reactors (SFRs), the phenomenology of sodium boiling during a postulated unprotected loss of flow (ULOF) transient has been investigated with the CATHARE 2 system code. This study focuses on a stabilized boiling case: in such a regime, no flow redistribution occurs from the subassemblies which have reached the saturation temperature to those that are still single-phase. In this paper, for a subassembly design featuring no restrictive structures above the fuel bundle, a quasi-static approach is first developed to get an upper bound of the reactor core power at boiling onset that would be compatible with the well-known Ledinegg criteria for diphasic flow static equilibrium. Then, dynamics results achieved through simulation with the CATHARE 2 code for a postulated ULOF are presented: boiling is shown to remain stable during the transient for such a core power at boiling onset. Another important outcome of the simulation is the calculation of a dynamic instability, in the form of a two-phase hydrodynamic chugging phenomenon. The predicted phenomenology of this stabilized boiling case should be studied further in order to consider its dependency on the underlying closure laws and to eliminate the possibility of a numerical instability.