Single-phase convective heat transfer in rod bundles

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.     

A new method to predict Grid-To-Rod Fretting in a PWR fuel assembly inlet region

Grid-To-Rod Fretting (GTRF) is one of the main causes of leaking fuel in a Pressurized Water Reactor (PWR). GTRF is caused by grid-to-rod gap, secondary flow, and axial/lateral turbulence caused pressure fluctuations within the fuel assembly, which produces rod vibration and wear. The cross flow and vortex shedding phenomenon produce low frequency vibration forces on fuel rods. In some plants, leaking fuel has been detected at the fuel inlet region of fuel assembly designs that do not have Protective Grid (P-grid) which, in addition to providing debris protection, also provides lateral stability against vibration. In order to understand the root cause of the fuel leaks, a thorough investigation of the flow field at the fuel inlet region is required. Leaking fuel has also been detected in the fuel inlet region in transition cores. In the transitional core arrangement, there are different fuel assembly designs next to each other. Due to the structure difference, there will be cross flow between fuel assemblies, which may be the initiating factor for fuel leaks.A method based on Computational Fluid Dynamics (CFD) has been developed in Westinghouse to predict the GTRF in the fuel inlet region. The fuel inlet region consists of the lower core plate, the bottom nozzle, the fuel rods, the thimble rods, the P-grid, and the bottom grid. This study employed CFD to investigate the unsteady forces on the fuel rods under typical reactor in-core conditions. Two fuel assembly (FA) inlet regions with and without the P-grid were simulated. The time history of the unsteady force components on fuel rods was recorded. Fast Fourier Transform (FFT) analyses were carried out for the force history. Compared to the data from operating plants, the new method predicted synchronized excitation forces on the rods that leaked in real operation. The CFD results also demonstrated the advantage of using the P-grid. GTRF at the fuel inlet region can be significantly reduced when the P-grid is used in Westinghouse fuel assembly designs.     

有界域轴向流动中棒束流致振动和稳定性研究

基于Ｃｈｅｎ有界域静止流体中棒束流致振动数学模型,给出了棒束稳定分析的数值方法。作为应用研究,分析了有界域轴向流动中棒束（４根）流致振动的特性及稳定性,给出了不同条件下轴向及横向的振型。     

模拟燃料棒通电加热的可视化实验研究

长直棒束是一种结构简单的电加热模拟燃料棒组件。在通电条件下,各单棒间将产生一定的相吸相斥效应。为保证这种加热组件使用的安全性,本实验利用高速摄像机对由9棒束中各棒通电加热后产生的相吸相斥现象进行研究。可视化实验研究结果表明,棒束通入直流电压后将产生一定程度的偏移。通过高速摄像机的逐频播放,观察到当通电电压增加到85V、通电电流为450A时,棒与棒间最大偏移距离约为1mm,且各单棒在通入直流电后的相吸相斥效应随直流电流的大小和通电方向变化而变化。棒束通入交流电后也产生相吸相斥效应,但其产生的效果没有通入直流电后产生的明显。     

压水堆燃料棒在轴向流作用下的随机振动响应研究

基于随机振动理论,建立了在轴向流作用下压水堆燃料棒随机响应的纯理论分析方法。将流体力考虑为沿燃料棒轴向位置的脉冲随机荷载,结合模态分析技术,从功率谱分析法推导出燃料棒振动均方根响应的表达式。提供了一套不依赖燃料组件流致振动实验的纯理论分析方法,重点分析了等效流速、湍流强度、相关长度系数等几个主要流场参数对结构均方根响应的影响。结果表明,本文计算模型的精度满足工程分析要求,燃料棒响应随等效流速、湍流强度和相关长度系数的增大而增大;其中响应对于等效流速和相关长度系数的变化较为敏感,而与湍流强度呈线性变化关系;在压水堆运行中的燃料棒均方根幅值约处在μm量级。     

New fuel rod design based on LEU/Th fuel and an assessment on in-core fuel cycle length of the MHR core

The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and depleted uranium in separate TRISO particles, in a single fuel rod within a graphite matrix. The TRISO particle volume packing fraction (PF) in the fuel rods is 29%, of which the LEU particle PF is 62%. The lifetime between refuelings is about 476 effective full power days (EFPD). In this paper we assess the possibility of replacing the depleted uranium TRISO particles with thorium TRISO particles, and evaluate the impact of such replacement on fuel cycle length. A preliminary scoping study was performed to determine the most promising fuel rod/zoning configurations. The scoping study indicates that there is advantage to separating the thorium TRISO particles from the LEU particles at the fuel rod level instead of mixing them within a single rod. An axial checkerboard distribution of the fuel rods where all uranium and all thorium rods are interchangeable along the axial direction within the graphite block is the most promising configuration that was identified in this study and can be lead to a fuel cycle length extension of 50-80% relative to the current design, with only a modest increase in the fissile material loading (15-20%). To this advantage can be added the benefit of a significant reduction in nuclear waste and in health risk. This study also lays the foundation for improving the fuel rod arrangement within the graphite block and the graphite blocks within the entire reactor core. The analysis is limited to a once - through fuel cycle based on in situ fissioning of the U-233, without further separation and reprocessing. The preliminary heat transfer analysis indicates that the maximum temperature in the fuel will be raised by about 10-15% over that of current MHR design.     

带格架5×5燃料棒束流致振动特性数值研究

核反应堆中,流动的冷却剂轴向冲刷燃料棒可能导致其振动,产生微动磨损,对整个核电厂的安全性以及经济性有重要影响。带格架棒束流致振动特性的研究是微动磨损研究的基础。本文基于欧拉-伯努利（Euler-Bernoulli）梁理论,采用动网格技术,通过Fluent实现流固耦合数值计算,并与不考虑振动耦合时的流场分布进行比较分析。重点分析了湍流强度、轴向速度等主要流体参数对振动位移均方根的影响,以及轴向流中流致振动机理。结果表明：燃料棒的振动位移均方根随着流速的增大而增大;燃料棒径向两侧的压力脉动是造成振动的因素之一;定位格架改变了较大振动出现的位置,明显加强了振动响应。     

燃料组件典型栅元湍流激振数值研究

为掌握全长范围内的燃料棒振动响应特性,以用于燃料棒微动磨损寿命分析,本研究运用计算流体动力学（CFD）方法,对燃料组件典型栅元的湍流激振进行数值模拟分析,并通过棒表面的瞬态脉动压力分布开展不同夹持力下的单棒瞬态动力学分析。研究表明:格架上游的截面平均湍动能约为0.1 m2/s2,格架临近出口位置湍动能达到峰值的0.65 m2/s2,格架的存在显著增强了流场的湍流强度,这是造成燃料棒湍流激振的主要原因;通过瞬态动力学分析确定了均方根振幅最大的定位格架位置,并建立了该格架的均方根振幅和振动速度随夹持力变化的关联式。本研究将为后续微动磨损理论计算及实验验证奠定基础。      

Thermal Resistance between Hexagonally Arrayed Channels

A new power-flattening method has been proposed for boiling water reactors (BWRs) which have an axially skewed power distribution caused by the void fraction distribution. In present BWRs, the skewed power distribution is avoided by using shallow control rods and/or axially distributed gadolinia fuel bundles. These means are effective for the axial power shape control, but perturb the self-power-flattening effect due to fuel burnup. The power-flattening method proposed here extensively utilizes this effect in the equilibrium cycle core. Based on this method, a new BWR core design and operating strategy, the WNS core concept, has been realized for reactor operation with no shallow control rod insertion and no fuel bundle shuffling. Studies of the WNS core has shown that the proposed power-flattening method has the potential to improve capacity factors, increase operating thermal margins and simplify reactor operations in comparison with current BWR cores.     

压水堆燃料辐照后燃料棒棒间距数据机器视觉测量方法研究

针对辐照后燃料棒棒间距数据获取和处理困难的问题,基于燃料棒几何特性及其在压水堆燃料组件中的排列方式,本文提出一种基于机器视觉的高效、可靠的燃料棒棒间距数据测量方法。该方法首先采用Retinex算法对水下燃料棒的采集图像进行增强预处理;然后,针对燃料棒阵列的前后成像干扰问题,采取边缘增强和逐行灰度特征处理方法实现待测燃料棒与背景燃料棒的有效分离,并进一步提升图像清晰度;最后,对燃料棒图像的单行灰度值进行二次曲线拟合,获得各个燃料棒的亚像素边缘点坐标。乏燃料组件的现场实验验证结果表明,该方法可一次性实现16个燃料棒棒间距测量,且测量精度达±0.32 mm,可为燃料性能分析提供高效、可靠的数据支持。      

CFD application to prediction of void distribution in two-phase bubbly flows in rod bundles

This paper is concerned with the prediction of the void fraction distribution in two-phase bubbly flows in fuel rod bundles. Special attention has been devoted to the phenomena which govern the void fraction distribution in the lateral direction of a channel. A two-fluid model of two-phase flow has been formulated and implemented into a commercial computational fluid dynamics (CFD) code. The model has been used for the prediction of the void distribution in three different channels: a circular channel (inside diameter (ID), 34.5 mm) with a single heated rod of 13.9 mm outside diameter (OD), and circular channels (ID, 71 mm) with six heated rods (13.8 and 13.9 mm OD each). The predicted axial and lateral avoid fraction distributions in subcooled and bulk boiling regions have been area averaged in three lateral zones and compared with experimental data: in all cases, satisfactory agreement between the predictions and measurements has been obtained.     

Testing of Lwr Fuel Rods to Support Criticality Safety Analysis of Transport Accident Conditions

Abstract

For the transport of low enriched materials, criticality safety may be emonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where light water reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data are needed. This requires a method by which the response of LWR fuel under accident impact conditions can be approximated or bounded. In 2000, British Nuclear Fuels and Areva Cogema Logistics jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. ACL were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the transport regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and ACL would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasistatic loading on fuel samples. Pressurised water reactor (PWR) fuel rods loaded with uranium pellets were dropped vertically from 9 m onto a rigid target and this was repeated on boiling water reactor (BWR) fuel rods; similar tests on empty fuel rods were also conducted. Quasistatic tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high-burn-up fuel rods of both PWR and BWR types. These were believed to be original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500°C during loading. All specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data were gained from this test programme.     

冷却剂与含格架燃料棒束多场耦合分析

燃料棒束作为压水堆燃料组件的组成部分,其热工和结构特性直接关系到反应堆的安全。本文利用ANSYS WORKBENCH软件分析了冷却剂在5×5含定位格架燃料棒束通道内流动的分布,采用冷却剂与燃料棒束多场耦合的方式研究了燃料棒束的流动传热特性和结构形变特性。结果表明：定位格架扰动冷却剂形成横向二次流并在下游棒束间形成绕流;多场耦合条件下二次流峰值速度和平均速度均小于单流场的;二次流与燃料棒的热应力使棒束发生形变,功率和流动分布的不均匀导致形变在轴向和径向的不均匀;相较于无格架情况,定位格架的存在使冷却剂的搅混流动更加明显,冷却剂对燃料棒冲击增大;在有、无定位格架两种情况下棒束形变均很小,可保持原本结构的稳定。     

Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code.The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from the plenum - resulting in a considerable delay of the predicted moment of cladding rupture.     

Local thermal–hydraulic behaviour in tight 7-rod bundles

Advanced water-cooled reactor concepts with tight lattices have been proposed worldwide to improve the fuel utilization and the economic competitiveness. In the present work, experimental investigations were performed on thermal–hydraulic behaviour in tight hexagonal 7-rod bundles under both single-phase and two-phase conditions. Freon-12 was used as working fluid due to its convenient operating parameters. Tests were carried out under both single-phase and two-phase flow conditions. Rod surface temperatures are measured at a fixed axial elevation and in various circumferential positions. Test data with different radial power distributions are analyzed. Measured surface temperatures of unheated rods are used for the assessment of and comparison with numerical codes.In addition, numerical simulation using sub-channel analysis code MATRA and the computational fluid dynamics (CFD) code ANSYS-10 is carried out to understand the experimental data and to assess the validity of these codes in the prediction of flow and heat transfer behaviour in tight rod bundle geometries. Numerical results are compared with experimental data. A good agreement between the measured temperatures on the unheated rod surface and the CFD calculation is obtained. Both sub-channel analysis and CFD calculation indicates that the turbulent mixing in the tight rod bundle is significantly stronger than that computed with a well established correlation.     

燃料组件导向管冷壁效应对CHF影响试验研究

燃料组件临界热流密度（CHF）性能是压水堆堆芯热工水力设计和安全分析的基础,对反应堆的安全运行至关重要。本文采用非均匀加热典型栅元和导向管栅元两组CHF试验数据开发了具有针对性的CHF关系式并对比研究了导向管冷壁对CHF的影响,获得了导向管冷壁效应因子。研究结果表明,针对轴向功率非均匀加热,在边界条件一致的情况下导向管的存在不会降低棒束的平均功率,但会导致烧毁点的位置发生变化并使得CHF降低,导向管冷壁效应因子约为8%。     

Experiment and numerical simulation of bubbly two-phase flow across horizontal and inclined rod bundles

Experimental and numerical analyses were carried out on vertically upward air-water bubbly two-phase flow behavior in both horizontal and inclined rod bundles with either in-line or staggered array. The inclination angle of the rod bundle varied from 0 to 60° with respect to the horizontal. The measured phase distributions indicated non-uniform characteristics, particularly in the direction of the rod axis when the rods were inclined. The mechanisms for this non-uniform phase distribution is supposed to be due to: (1) Bubble segregation phenomenon which depends on the bubble size and shape; (2) bubble entrainment by the large scale secondary flow induced by the pressure gradient in the horizontal direction which crosses the rod bundle; (3) effects of bubble entrapment by vortices generated in the wake behind the rods which travel upward along the rod axis; and (4) effect of bubble entrainment by local flows sliding up along the front surface of the rods. The liquid velocity and turbulence distributions were also measured and discussed. In these speculations, the mechanisms for bubble bouncing at the curved rod surface and turbulence production induced by a bubble were discussed, based on visual observations. Finally, the bubble behaviors in vertically upward bubbly two-phase flow across horizontal rod bundle were analyzed based on a particle tracking method (one-way coupling). The predicted bubble trajectories clearly indicated the bubble entrapment by vortices in the wake region.     

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis.     

不同约束条件和几何参数对有界域轴向流动棒束稳定性的影响

基于有界域轴向流动中棒束流致振动数学模型和棒束稳定性的分析方法,作为应用研究,分析了棒束端约束条件和系统几何参数对系统失稳时临界流速的影响。