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同一软件工具采用不同湍流模型进行燃料组件格架棒束通道CFD分析时会得到不同的数值结果,本文采用ANSYS CFX软件,建立了包含典型5×5格架的棒束通道CFD模型,研究了涡粘和雷诺应力两大类6种典型湍流模型对燃料组件压降与换热特性数值结果的影响,计算了压降和Nu分布结果与相似的实验结果进行对比,通过分析3个典型搅混效果评价因子,探讨了搅混翼流动与换热的内在影响关系,同时对比了不同湍流模型对结果的影响。通过与相似实验数据对比分析,认为雷诺应力模型较适宜计算本文所研究的定位格架及棒束通道内流动传热特性。 相似文献
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为了研制高性能燃料组件,定位格架的阻力特性直接关系到燃料组件的热工性能和水力相容性。本文针对5×5规模的定位格架,从流动阻力的基本原理出发,利用CFD方法研究并建立了格架局部阻力特性的理论计算模型,并对计算结果进行验证。结果表明,基于计算模型获得的格架局部阻力系数与直接模拟结果基本一致,验证了计算模型的准确性。 相似文献
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定位格架作为燃料组件的关键部件之一,直接影响到燃料组件的热工性能。本文对带结构格架(MVG)和跨间搅混格架(MSMG)的5×5全长加热棒束单相流场和温度场采用计算流体力学(CFD)程序进行数值分析研究,获得该特征棒束组件出口二次流场以及温度场分布特性。研究表明,定位格架下游流场受定位格架和距离的影响,定位格架上游流场对下游二次流几乎无影响,定位格架导致流体强烈的横向二次流,增强了流体和加热棒之间的换热能力,使得棒束子通道截面流体温度更加均匀。与5×5全长棒束出口子通道温度的实验数据对比分析表明,获得的计算模型可以较好地分析该型棒束组件结构温度场行为。 相似文献
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采用两相计算流体动力学(CFD)方法进行带7道格架的5×5棒束两相性能研究,其中结构搅混格架(MG)和跨间搅混格架(MSMG)交替布置,计算考虑汽泡合并与破裂、热量传递,但不考虑相间的质量传递。为选择合理的两相模型参数,首先以带2道格架(MG、MSMG)的AFA3G燃料组件5×5棒束架为研究对象,对最大气泡直径、汽泡合并破裂系数、非曳力模型及曳力模型、入口气泡直径、入口空泡份额分布等进行了敏感性及不确定性分析。此后采用该两相模型设置,针对带7道格架的AFA3G燃料组件进行了两相性能研究,计算结果显示格架间的各项参数不存在完全一致的周期性,但同种格架上游的空泡份额分布具有一定的相似性,因此用于两相性能评价可计算带2~3道格架的棒束,该研究可用于带格架棒束两相计算的模型设置与几何规模选择,为下一步采用两相CFD计算建立燃料组件热工水力性能评价准则奠定了基础。最后比较了AFA3G燃料组件及改进型燃料组件两种格架的空泡分布特性,并从提高燃料组件临界热流密度(CHF)特性的角度对其进行评价,获得与实验一致的结论,证明了评价方法的正确性。 相似文献
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计算流体动力学在燃料组件热工水力性能分析和格架研制中的应用 总被引:4,自引:1,他引:3
国外使用商用计算流体动力学(CFD)软件分析燃料组件中流体的三维流场和温度场,并将验证的方法用于燃料组件格架设计,获得了成功。中国核动力研究设计院空泡物理和自然循环重点实验室用CFX程序对带格架棒束内流场进行了计算,解决了小尺寸复杂结构几何体的模拟,边界条件的选取和CFX计算能力的评价,然后完成了单相,空气一水两相流场和流动特性的计算分析及试验对比验证。已完成的研究表明,尽管CFX程序目前在计算两相流动和传热方面还存在不足,但通过比较单相流场的湍流,旋涡和棒束附近流体温度分布基本可以评价格架对流体的交混性能;格架上的弹簧和刚突对于流动有相当的作用,对其进行模拟是必要的。研究还建议在使用CFD方法进行燃料组件格架热工水力分析前要先进行基准练习以保证分析结果的正确性。 相似文献
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燃料组件格架几何建模及网格划分技术 总被引:2,自引:0,他引:2
为采用计算流体力学(CFD)方法对燃料组件格架的搅混性能进行研究,对燃料组件格架几何建模及网格划分进行了系统的研究。比较不同几何模型得到的计算结果,确定了将搅混格架简化为无刚凸、无弹簧、只有条带和搅混翼结构的模型;出、入口段的模拟长度分别为250mm和230mm;模拟格架的数量为一道。研究网格对计算结果的影响,确定了分段网格划分方式和网格数量:入口段和出口段采用结构化网格,格架段采用非结构化网格。整个研究对象总节点数为1032258,总栅元数为2601614,其中格架段网格数占75.8%。 相似文献
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提出了一种基于NHR200-II供热堆燃料组件定位格架的简化模型。简化建模方法包括2方面:将定位格架上的内刚凸及三弯弹簧用非线性连接器代替;使用梁单元代替实际燃料棒。结合前期关于NHR200-II定位格架的研究成果,确定了非线性连接器的刚度,并通过有限元软件建立了燃料组件简化前后的1×2局部子模型,分析了其固有频率与碰撞特性,证明了简化建模方法的有效性。随后,该简化方法被应用于全尺寸的9×9定位格架模型,研究了格架夹持能力对动力学特性的影响,结果表明,该简化方法可以有效地模拟不同夹紧程度下格架的地震谱响应。综上,从有限元建模角度来看,本文提出的基于NHR200-II燃料组件定位格架的方法是有效的。 相似文献
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分别以实验与数值模拟对5×5棒束通道压降特性进行了研究。在5×5棒束通道实验本体上开展了压降实验研究,雷诺数范围为2000~14000。获得了棒束通道内压降随雷诺数的变化关系,并在实验工况范围内拟合了摩擦阻力系数计算经验关系式,关系式对摩擦阻力系数的预测偏差在5%以内。在实验研究基础上,开展了棒束通道内压降数值研究。对于雷诺数低于2000的工况选取层流模型,雷诺数高于2000的工况选取标准k-ε模型、Realized k-ε模型、RNG k-ε模型与LPS-RSM等湍流模型,开展了棒束通道内流场数值模拟,并拟合了层流工况下高精度摩擦阻力系数计算关系式。数值模拟结果表明,雷诺数较高时,标准k-ε模型、Realizedk-ε模型、RNG k-ε模型与LPS-RSM等湍流模型均能较好地预测摩擦阻力特性。 相似文献
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The present paper discusses entropy generation in fully developed turbulent flows through a subchannel,arranged in square and triangle arrays. Entropy generation is due to contribution of both heat transfer and pressure drop. Our main objective is to study the effect of key parameters such as spacer grid, fuel rod power distribution,Reynolds number Re, dimensionless heat power ω, lengthto-fuel-diameter ratio λ, and pitch-to-diameter ratio ξ on subchannel entropy generation. The analysis explicitly shows the contribution of heat transfer and pressure drop to the total entropy generation. An analytical formulation is introduced to total entropy generation for situations with uniform and sinusoidal rod power distribution. It is concluded that power distribution affects entropy generation.A smoother power profile leads to less entropy generation.The entropy generation of square rod array bundles is more efficient than that of triangular rod arrays, and spacer grids generate more entropy. 相似文献
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Mary V. Holloway Donald E. Beasley Michael E. Conner 《Nuclear Engineering and Design》2008,238(4):848-858
The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids. 相似文献
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The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due to the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5 × 5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 sub-domains. The sub-domains were simulated sequentially applying the outlet results of a previous sub-domain as inlet condition for the next. In this study the k-ε turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and In et al. in one subchannel computational domains. Comparison between numerical and experimental results of lateral and axial velocities along of the rod bundle show good agreement for all evaluated heights downstream the spacer grid. The present numerical procedure shows better predictions than Karoutas et al. model especially further from the spacer grid where the peripheral subchannels have more influence in the average flow. 相似文献
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《Progress in Nuclear Energy》2012,54(8):1190-1196
The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due to the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5 × 5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 sub-domains. The sub-domains were simulated sequentially applying the outlet results of a previous sub-domain as inlet condition for the next. In this study the k–ε turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and In et al. in one subchannel computational domains. Comparison between numerical and experimental results of lateral and axial velocities along of the rod bundle show good agreement for all evaluated heights downstream the spacer grid. The present numerical procedure shows better predictions than Karoutas et al. model especially further from the spacer grid where the peripheral subchannels have more influence in the average flow. 相似文献
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以中国超临界水冷堆(CSR1000)燃料组件研发为研究背景,采用实验辅以理论分析的方法,开展2×2棒束结构内超临界水工质的传热特性研究。实验工况范围为:压力(P)23~25 MPa;质量流速(G)680~1400 kg/(m2?s);热流密度(q)174~968 kW/m2。实验结果表明,随着q的增加、G的减小,2×2棒束的传热性能减弱;随着P从23 MPa变化到25 MPa,2×2棒束的传热性能变化微弱; 2×2棒束内超临界水的传热特性既与边界层和主流的物性差异程度有关,又受流道各子通道之间的流动传热不均匀性影响;基于实验数据进行多元线性回归分析,获得2×2棒束内超临界水换热关系式,约88.9%的实验数据与该换热关系式的计算值偏差范围在±25%内。 相似文献