基于丝网传感器的5×5棒束通道空泡分布测量研究

为研究压水反应堆燃料组件棒束通道内的两相分布规律,设计并制造了适用于棒束通道的丝网传感器模块,开展了5×5棒束通道内空气-水泡状流的空泡分布测量实验,分析了棒束通道内空泡份额的分布规律及气泡尺寸对空泡分布的影响。实验结果表明,发生横升力方向反转的小气泡在壁面附近聚集、大尺寸气泡则聚集在子通道中心;常温常压下发生横升力方向反转的临界气泡直径在4~6 mm之间,证明了横升力模型在棒束通道中的适用性。      

棒束定位格架两相CFD模拟方法研究

考虑气泡合并分裂,采用MUSIG模型,对3×3格架内空气-水两相分布进行计算流体力学(CFD)数值模拟研究发现,计算对入口两相分布预计不敏感,但对气泡直径大小敏感;在定位格架下游不远处,空泡份额分布由较小直径气泡起主导作用,格架下游较远处,空泡份额分布由较大直径气泡起主导作用。考虑空气-水两相流量、几何条件和压力对气泡直径的影响,本文提出针对棒束定位格架的数值模拟气泡最大直径设置关系式,并对模型选取和模拟方法给出建议。计算表明空泡份额分布曲线形状与峰值均和实验符合较好,该模拟方法能合理预测复杂通道两相数值分布。     

竖直圆管内泡状流空泡份额径向分布实验研究

常温常压下,采用光学探针测量方法,对圆管（内径50 mm）内空气 水两相竖直向上泡状流空泡份额的径向分布特性进行了实验研究。结果表明,竖直圆管内泡状流空泡份额的径向分布随气液两相表观流速不同而变化。液相流速较高时空泡份额分布呈“壁峰型”,即中心区域变化平缓,近壁区出现峰值后迅速降低;液相静止时,随气相流速增加,空泡份额增加速度沿径向向外逐渐减小,气相流速较大时分布呈“核峰型”,即空泡份额随径向位置向外呈减小趋势;液相流速较低时分布呈现出过渡型。探针测量面积加权平均空泡份额与通过重位压降得到的空泡份额的相对偏差小于10%。     

矩形通道中两相流变密度模型空泡份额计算

基于Bankoff的圆管内和无限长平板间两相流变密度模型空泡份额计算式的推导，结合流体在管道中的流场分布特征，建立了矩形通道中两相流流场分布规律方程，导出了变密度模型在矩形通道中空泡份额的计算式，并对3种通道计算的结果进行了对比分析。计算结果与原有Bankoff模型符合得很好。     

Experimental and numerical simulation of air-water two-phase flow in the rod bundle with grid spacer

1 Introduction Grid spacer is the key part of reactor fuel assem-bly. The presence of spacers in fuel assemblies affectsvarious thermal-hydraulic characteristics of the reactorcore. The grid spacer with fine performance can im-prove thermal-hydraulic performance of the core fuelassembly and enhance the critical heat flux withouttoo much augment of the pressure loss. As a result,the implementation of grid spacer with high thermalperformance provides more thermal margin, then in-creases s…     

Three-dimensional time-averaged void fraction distribution measurement technique for BWR thermal hydraulic conditions using an X-ray CT system

We have developed a void fraction distribution measurement technique using the three-dimensional (3D) time-averaged X-ray computed tomography (CT) system to understand two-phase flow behavior inside a fuel bundle for boiling water reactor (BWR) thermal hydraulic conditions of 7.2 MPa and 288 °C. As a first step, we measured the 3D void fraction distribution in a vertical square (5?×?5) rod array that simulated a BWR fuel bundle in the air–water test. A comparison of the volume-averaged void fractions evaluated by the developed X-ray CT system with those evaluated by a differential pressure transducer showed satisfactory agreement within a difference of 0.03. Thus, we confirmed that the developed system could be used to get 3D imaging of the vertical square rod array used in the test under the BWR operating pressure condition. In the next step, we did a verification test using the vertical pipe (11.3 mm ID) for BWR thermal hydraulic conditions. A comparison of the cross-sectional-averaged void fractions evaluated by the X-ray CT system with those evaluated by the drift-flux model showed good agreement within a difference of 0.05. We confirmed that the evaluated void fraction distribution forms in the horizontal cross section changed with the quality in response to the flow regime transition.     

Modeling of distribution parameter,void fraction covariance and relative velocity covariance for upward steam-water boiling flow in vertical rod bundle

In order to improve the prediction accuracy of one-dimensional interfacial force formulated by ‘Andersen’ approach, the distribution parameter in a drift–flux correlation, void fraction covariance, and relative velocity covariance has been modeled for dispersed boiling two-phase flow in a vertical rod bundle. The distribution parameter has been derived by a bubble-layer thickness model. The correlations of void fraction covariance and relative velocity covariance have been developed based on prototypic 8 × 8 rod bundle data. The correlation of void fraction covariance agrees with the bundle data with the mean absolute error, standard deviation, mean relative deviation, and mean absolute relative deviation being 0.00120, 0.0415, ?0.173%, and 1.80%, respectively. The correlation of relative velocity covariance agrees with the bundle data with the mean absolute error, standard deviation, mean relative deviation, and mean absolute relative deviation being ?0.00241, 0.0452, ?0.0316%, and 2.52%, respectively. In view of the great importance of void fraction covariance and relative velocity covariance on the one-dimensional interfacial drag force formulation, it is highly recommended to include the void fraction covariance and relative velocity covariance in the one-dimensional formulation of interfacial drag force used in nuclear thermal-hydraulic system analysis codes.     

Critical Heat Flux Prediction Method Based on Two-Phase Turbulence Model

In this paper, we present an analytical methodology to predict forced convective CHF (Critical Heat Flux) for DNB (Departure from Nucleate Boiling) type boiling transition that occurs inside of uniformly heated round tubes. Axial directional two-phase flow analysis was conducted based on one-dimensional two-fluid model and typical constitutive models. At the same time, the radial directional distribution of void fraction at any axial location was calculated based on the bubble diffusion model, which was coupled with two-phase turbulence model for boiling bubbly flow. The calculated void fraction showed the wall peak distribution, and was compared with experimental data, which was derived from subcool boiling experiments. IPNVG (Incipient Point of Net Vapor Generation), which means the starting point of two-phase flow analysis, was also investigated well, since it was revealed that IPNVG had a significant influence on CHF prediction. By using this methodology for calculating radial directional void fraction distribution, we carried out CHF prediction for water on the assumption that DNB would occur when the local void fraction near the heated wall exceeds a critical value. The predicted CHF agreed well with experimental data, and the accuracy was within about 20%.     

Analysis of subcooled boiling flow with one-group interfacial area transport equation and bubble lift-off model

To enhance the multi-dimensional analysis capability for a subcooled boiling two-phase flow, the one-group interfacial area transport equation was improved with a source term for the bubble lift-off. It included the bubble lift-off diameter model and the lift-off frequency reduction factor model. The bubble lift-off diameter model took into account the bubble's sliding on a heated wall after its departure from a nucleate site, and the lift-off frequency reduction factor was derived by considering the coalescences of the sliding bubbles. To implement the model, EAGLE (elaborated analysis of gas-liquid evolution) code was developed for a multi-dimensional analysis of two-phase flow. The developed model and EAGLE code were validated with the experimental data of SUBO (subcooled boiling) and SNU (Seoul National University) test, where the subcooled boiling phenomena in a vertical annulus channel were observed. Locally measured two-phase flow parameters included a void fraction, interfacial area concentration, and bubble velocity. The results of the computational analysis revealed that the interfacial area transport equation with the bubble lift-off model showed a good agreement with the experimental results of SUBO and SNU. It demonstrates that the source term for the wall nucleation by considering a bubble sliding and lift-off mechanism enhanced the prediction capability for the multi-dimensional behavior of void fraction or interfacial area concentration in the subcooled boiling flow. From the point of view of the bubble velocity, the modeling of an increased turbulence induced by boiling bubbles at the heated wall enhanced the prediction capability of the code.     

空气-水在垂直非圆截面通道内流型研究

采用可视化方法研究了水力直径分别为15mm和10mm的两种正方形截面、14.43mm的三角形截面以及14mm的圆形截面通道内空气-水垂直上升流动,表观气速0.04～80m/s,表观水速0.001～6m/s.观察到了泡状流、弹状流、块状流、环状流和弥散泡状流等常见流型.此外,在表观气速很大而表观水速很小时,在非圆截面通道内发现了爬动流,证实了非圆截面直通道内存在"二次流"现象,且对气-液两相流动的相分布有较大影响,证明截面形状对两相流流型及其转变具有重要影响.由实验得到了流型转变界限,并首次获得了包括爬动流的两相流流型图.比较本文的实验结果及与前人的研究结果对比发现,水力直径的大小对两相流流型的转变具有一定影响.