基于流热固耦合的核电蒸汽发生器传热管热应力数值模拟

以大亚湾核电站蒸汽发生器为原型,基于相似模化原理建立了蒸汽发生器简化物理模型。采用两流体模型及热弹性力学基本关系式分别描述气液两相流沸腾相变过程和热应力变化规律。利用CFX对一、二回路侧流体流动传热及与传热管的耦合换热过程进行了数值模拟,并在ANSYS WORKBENCH中实现了流体温度场载荷向结构的传递,进而对传热管进行稳态热分析和热应力分析。计算结果表明：二回路出口质量含汽率为24.5%,冷却剂出口温度为296.2 ℃,均与大亚湾蒸汽发生器实际运行参数相符;传热管热应力与其壁面温差分布一致,且沿壁厚方向先减小后增大,并存在中性层,传热管最大热应力为54.5 MPa。研究结果为蒸汽发生器的优化设计及安全运行提供了一定的理论支撑。     

不同运行工况下蒸汽发生器热工水力稳态特性数值研究

基于相似模化理论建立了蒸汽发生器一、二回路流体及传热管流 固耦合传热的单元管三维物理模型,对大亚湾核电厂蒸汽发生器不同工况下的热工水力稳态特性进行了数值模拟研究。采用热相变模型描述二回路汽液两相流动与换热、流-固耦合模型描述一回路冷却剂借助U型管与二回路流体换热。数值计算结果表明：满负荷运行时,传热管内壁温度变化趋势与一次侧流体基本一致,外壁温度与二次侧流体温度变化趋势相同;截面平均含汽率沿传热管高度的升高呈上升趋势,出口质量含汽率与大亚湾核电厂实际运行参数相符;随负荷降低一回路出口温度基本不变,二回路出口温度升高,质量含汽率及传热系数下降,平均传热系数与Rohsenow经验关联式的计算结果基本吻合。     

蒸汽发生器二次侧汽液两相流数值模拟

以大亚湾核电站蒸汽发生器为原型,在相似原理的指导下,建立了蒸汽发生器“单元管”三维物理模型,采用Particle模型和热力学相变模型,并基于CFX软件实现了蒸汽发生器二回路侧两相流流动与沸腾换热特性数值模拟。计算结果表明：满负荷运行时,沿传热管高度升高,蒸汽发生器的传热系数及截面含汽率均呈上升趋势,其平均传热系数的数值模拟结果与Rohsenow经验关联式计算结果间的误差为8.4%,出口质量含汽率与大亚湾核电站实际运行参数相符。热相变模型在蒸汽发生器两相流数值模拟中的成功应用,可为蒸汽发生器热工水力的准确分析提供参考。     

自然循环系统摇摆条件下棒束通道内传热特性研究

棒束燃料元件子通道间流体存在搅混与横向二次流,流动及阻力特性相较矩形通道、圆管等简单通道更为复杂。核动力舰船、船舶、小型浮动核电站等会受到海浪影响,经常处于倾斜、摇摆、垂荡等瞬变运动下。目前的相关研究多集中在低压工况的研究领域,高温高压自然循环运动条件下的研究较少。本文采用实验研究方法,对自然循环系统摇摆条件下棒束通道内流动传热特性进行了研究,获得了过冷沸腾和饱和沸腾两种条件下摇摆角度和摇摆周期对棒束壁面温度变化和传热系数的影响,并获得了摇摆周期内棒束通道内的传热系数计算关系式。结果表明,饱和沸腾传热系数变化比过冷沸腾的剧烈;在本文实验工况范围内,棒表面传热系数波动幅值随着摇摆幅度的增大而增大;摇摆条件下棒束通道过冷沸腾和饱和沸腾工况时均传热系数基本不变。     

Heat transfer in two-component internal mist cooling: Part II. Mechanistic modeling

The prediction of a mechanistic, three-dimensional, two-phase flow model is compared with experimental heat transfer data presented in the experimental part of this study for steady, internal, nozzle-generated, gas/liquid mist flow in vertical channels. The mechanistic model is based on the modification of the KIVA-3V computer code. The KIVA-3V code has been modified to solve the heat conduction equation in the surrounding structure with either steady or pulsed heat generation simultaneously with the fluid transport equations, and allow modeling of the various channel geometries and droplet injection methods. Among the numerically examined operating and design parameters are: the liquid atomization nozzle design, heat flux, carrier gas velocity and inlet temperature, liquid mass fraction at inlet, and flow direction. Comparison is made between the experimental data for wall and fluid bulk temperatures and heat transfer coefficients, and the predictions of the numerical model. Overall, reasonable agreement is obtained for downward mist flow, in particular at moderate heat fluxes; at high heat fluxes, the model slightly underpredicts the local heat transfer coefficients. For upward mist flow, the model underpredicts the local heat transfer coefficients typically by about 20%, and appears to predict dryout at the test section exit earlier than experiment. Some parametric and sensitivity calculation results are also presented and discussed.     

棒束子通道湍流交混特性的数值模拟研究

为了提高核反应堆系统的经济性和安全性,本文采用CFD方法对棒束子通道间湍流交混效应进行研究。对子通道建模,选取SST k-ω模型进行计算,完成了网格敏感性分析。采用类比浓度计算法与间隙湍流热流法对湍流交混系数进行计算。计算结果表明：雷诺数较小时,单相湍流交混系数随雷诺数的增大而增大;当雷诺数达到一定值时,单相湍流交混系数近似为定值;采用类比浓度计算法与间隙湍流热流法计算所得的湍流交混系数无太大差别。本文拟合得到了适用于单相工况的湍流交混系数计算公式。     

运行条件对蒸汽发生器热工参数影响的仿真研究

以减轻蒸汽发生器破管事故及考察核电站电力升级为目的,参考大亚湾核电站蒸汽发生器的运行参数,基于分布参数法建立了核动力蒸汽发生器一维数学模型,开发了基于MATLAB的动态仿真程序,进行了改变运行条件时蒸汽发生器热工参数仿真计算。计算结果表明:与满负荷正常运行条件相比,在降低二回路运行温度或增加二回路流量时,二回路预热段变短,出口焓大幅升高;质量含汽率在降低温度时提高54%,增加流量时提高28%;一、二回路及管壁整体温度降低;一回路和内壁温降增大。该计算结果揭示了蒸汽发生器的内在传热规律,可为缓解U形管恶化及提升电力的相关操作提供一定理论依据。     

乏燃料干式储存模块衰变热导出的三维仿真计算分析

建立了QM-400储存模块内自然对流换热的三维流动及换热的计算模型,计算了QM-400储存模块内部的三维流场和温度场,得到了压力分布和速度分布以及储存筒表面的温度分布.在储存筒内部,建立了三维流场和换热的计算模型,并进行了计算,获得了乏燃料包壳的表面温度.     

Numerical analysis of thermal striping induced high cycle thermal fatigue in a mixing tee

Thermal striping, characterized by turbulent mixing of two flow streams of different temperatures that result in temperature fluctuations of coolant near the pipe wall, is one of the main causes of thermal fatigue failure. Coolant temperature oscillations due to thermal striping are on the order of several Hz. Thermal striping high-cycle thermal fatigue that occurs at tee junctions is one of the topics that should be addressed for the life management of light water reactor (LWR) piping systems. This study focuses on numerical analyses of the temperature fluctuations and structural response of coolant piping at a mixing tee. The coolant temperature fluctuations are obtained from Large Eddy Simulations that are validated by experimental data. For the thermal stress fatigue analysis, a model is developed to identify the relative importance of various parameters affecting fatigue-cracking failure. This study shows that the temperature difference between the hot and cold fluids of a tee junction and the enhanced heat transfer coefficient due to turbulent mixing are the dominant factors of thermal fatigue failure of a tee junction.     

A heat, mass, and momentum transport model for hydrogen diffusion flames in nuclear reactor containments

It is now possible to analyze the time-dependent, fully three-dimensional behavior of hydrogen combustion in nuclear reactor containments. This analysis involves coupling the full Navier-Stokes equations with multi-species transport to the global chemical kinetics of hydrogen combustion. A transport equation for the subgrid scale turbulent kinetic energy density is solved to produce the time and space dependent turbulent transport coefficients. The heat transfer coefficient governing the exchange of heat between fluid computational cells adjacent to wall cells is calculated by a modified Reynolds analogy formulation. The analysis of a MARK-III containment indicates very complex flow patterns that greatly influence fluid and wall temperatures and heat fluxes.     

多轴疲劳模型对比研究及其在核电结构热疲劳寿命预测中的应用

压水堆核电站余热排出系统冷热水混合区管道发现的热疲劳问题影响核反应堆的安全。本文通过一种采用单轴疲劳试验数据拟合疲劳寿命曲线，进而用于预测多轴疲劳寿命的分析方法，基于文献中的疲劳试验数据，对Dang Van模型、Matake模型和Fatemi-Socie模型进行了余热排出系统冷热水混合区管道材料304L不锈钢疲劳寿命预测结果的对比研究。基于余热排出系统冷热水混合区管道的三维简化有限元模型，分别应用Dang Van模型、Matake模型和Fatemi-Socie模型对管道热疲劳寿命进行了预测，并与试验结果进行了对比验证。研究结果表明，基于应变（含平均应力修正）的Fatemi-Socie模型比较适用于304L不锈钢的疲劳分析，其热疲劳寿命预测结果相对Dang Van模型、Matake模型较合理。     

蒸汽发生器一维均相流模型及其换热性能

以大亚湾核电站蒸汽发生器为原型,在合理假设基础上,分别以一回路流体、二回路流体及U形管内、外壁为研究对象,建立了蒸汽发生器一维均相流模型。采用Jacobi迭代法编制了基于MATLAB的仿真程序,对蒸汽发生器不同工况下的稳态换热性能进行计算,得出各流程关键参数的变化规律,计算结果与大亚湾核电站实际运行数据相吻合。计算结果揭示了蒸汽发生器的内在传热规律,可为蒸汽发生器优化设计提供一定的理论依据。     

Comparison of thermo-hydraulic performances of large scale vortex flow (LSVF) and small scale vortex flow (SSVF) mixing vanes in 17 × 17 nuclear rod bundle

Spacer grids in the nuclear fuel rod assembly maintain a constant distance between rods, secure flow passage and prevent the damage of the rod bundle from flow-induced vibration. The mixing vanes attached to the spacer grids generate vortex flows in the subchannels and enhance the heat transfer performance of the rod bundle. Various types of mixing vanes have been developed to produce cross flows between subchannels as well as vortex flows in the subchannels.The shapes of the mixing vane have been improved to generate larger turbulence and cross flow mixing. In the present study, two types of large scale vortex flow (LSVF) mixing vanes and two types of small scale vortex flow (SSVF) mixing vanes are examined. SSVF-single is conventional split type and SSVF-couple is split type with different arraying method. LSVF mixing vane has different geometry and arraying method to make large scale vortex. 17 × 17 rod bundle with eight spans of mixing vanes is simulated using the IBM 690 supercomputer. The FLUENT code and IBM supercomputer is employed to calculate the flow field and heat transfer in the subchannels.Turbulence intensities, maximum surface temperatures of the rod bundle, heat transfer coefficients and pressure drops of the four kinds of mixing vanes are compared. LSVF mixing vanes produced higher turbulence intensity and heat transfer coefficient than SSVF mixing vanes. Consequently, LSVF mixing vane increases the thermal efficiency and safety of the rod bundle.     

稳压器波动管热分层应力及疲劳分析

稳压器波动管内流体的温度分层引起管壁温度分层,从而在管道截面产生整体弯曲应力、局部热应力以及管道系统超过预期的位移和支撑载荷.将稳压器波动管的热分层这种复杂的三维应力分析问题简化为一维和二维组合问题,利用SYSTUS程序和ROCOCO程序对秦山核电二期扩建工程稳压器波动管热分层的应力及疲劳进行了分析研究,计算了考虑热分...     

反应堆冷却剂系统主管道安注斜接管等温横向射流流动特性的研究

针对核反应堆冷却剂系统中的主管道安注斜接管等温横向射流问题,应用计算流体力学商用软件CFX进行等温横向射流流动的数值模拟,得出了典型运行工况下的三维流场分布.深入研究了射流与主流在不同流速比情况下等温横向射流的流动特性、影响区域及主要影响因素,在所研究的参数范围内,得出了流速比是决定等温横向射流流动特性的最重要因素,同时将数值模拟结果与流动可视化试验结果进行了比较,二者吻合良好.     

Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements

Considering the mutual actions between fuel particles and the metal matrix, the three-dimensional finite element models are developed to simulate the heat transfer behaviors of dispersion nuclear fuel plates. The research results indicate that the temperatures of the fuel plate might rise more distinctly with considering the particle swelling and the degraded surface heat transfer coefficients with increasing burnup; the local heating phenomenon within the particles appears when their thermal conductivities are too low. With rise of the surface heat transfer coefficients, the temperatures within the fuel plate decrease; the temperatures of the fuel plate are sensitive to the variations of the heat transfer coefficients whose values are lower, but their effects are weakened and slight when the heat transfer coefficients increase and reach a certain extent. Increasing the heat generation rate leads to elevating the internal temperatures. The temperatures and the maximum temperature differences within the plate increase along with the particle volume fractions. The surface thermal flux goes up along with particle volume fractions and heat generation rates, but the effects of surface heat transfer coefficients are not evident.     

Natural Convection Heat Transfer in the Horizontal Dry Storage System for the LWR Spent Fuel Assemblies

A heat transfer and flow visualization experiment was conducted with a one-fifth scale model simulating a dry shielded canister (DSC) with 24 PWR spent fuel assemblies in order to elucidate the heat transfer characteristics and the velocity distribution for natural convection inside a DSC filled with air or water at atmospheric pressure. It was found that the average heat transfer coefficients were proportional to the one-fourth power of the Rayleigh number despite the complicated geometry inside the DSC. Flow patterns inside the DSC were visualized clearly through a digital image processing system. The velocity distributions inside the DSC were obtained quantitatively from the Particle Tracking Velocimetry. In comparison with the results of a two-dimentional thermal hydraulic analysis, computed flow patterns were similar to the experimental results and the computational temperature distributions on the sleeve surfaces agreed well with the experiments within 8%, except at the top point of the center gap. It was also found that the difference in the heat transfer coefficient was within 25% for air as the working fluid, while a satisfactory agreement was not obtained when water was the working fluid.     

压水堆子通道欠热沸腾数值验证及交混翼研究

为改善压水堆交混翼格架在欠热沸腾工况下的热工水力特性,以子通道为研究对象验证了所使用的欠热沸腾数值模型在不同工况下的有效性。基于已验证的数值模型,对含不同偏折角交混翼格架的子通道模型在不同工况下进行了两相流数值模拟,研究交混翼及其偏折角对子通道中两相流动、传热及气泡分布的影响。结果表明：交混翼在增大压降的同时明显强化了冷却剂的交混、降低了近壁面气泡份额、提高了换热效率,且在一定范围内偏折角越大影响越明显。相对较高的气泡份额将导致更大的压力损失、减弱冷却剂的交混、降低传热效率。当交混翼偏折角达25°时,继续增大其偏折角对降低近壁面气泡份额和提高传热效率的作用不再明显,反而造成压降的快速增大,因此建议其偏折角在25°左右。     

Heat Transfer by Slug Flow in Annulus

This paper deals with an analysis of heat transfer by slug flow in an annulus with different heat fluxes at the two wall surfaces. The local Nusselt numbers at each wall surface are calculated upon analyzing the radial distribution of the fluid temperature.

The local Nusselt number varies with the ratios of the radii as well as of the wall heat fluxes. The heat transfer in a circular tube and between two parallel plates are also discussed. As a special case, the present results agree very well with Dwyer's obtained for the heat transfer between two wall surfaces.     

蒸汽发生器二次侧两相流传热特性数值研究

以AP1000核电站蒸汽发生器为原型,建立蒸汽发生器二次侧"平均通道"模型,利用计算流体动力学软件ANSYS CFX,基于相界面模型对蒸汽发生器二次侧两相流流动和沸腾换热过程进行研究。结果表明:数值模拟计算方法能准确模拟蒸汽发生器二次侧汽液两相流沸腾和传热过程;满负荷运行时,流体由预热区经过泡核沸腾区过渡到稳定沸腾区,含汽率和传热系数沿流动方向逐渐增大,出口含汽率与该型号蒸汽发生器设计值符合较好,平均传热系数的模拟结果和JensLottes经验关联式的预测值基本一致。