Grid-to-rod flow-induced impact study for PWR fuel in reactor

Dynamic contact impact from hydraulic flow-induced fuel assembly vibration is the source for grid-to-rod fretting in a pressurized water nuclear reactor (PWR). To support grid-to-rod fretting wear mitigation research, finite element analysis (FEA) was used to evaluate the hydraulic flow-induced impact intensity between the fuel rods and the spacer grids. Three-dimensional FEA models, with detailed geometries of the dimple and spring of the actual spacer grids along with fuel rods, were developed for flow impact simulation. The grid-to-rod dynamic impact simulation provided insights of the contact phenomena at grid-rod interface. It is an essential and effective way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.     

棒束通道内定位格架搅混特性PIV可视化研究

定位格架作为燃料组件中重要的组成部件之一,不仅在结构上固定燃料棒,而且在燃料组件内热工水力性能同样显著,特别是对工质的搅混性能直接关系到反应堆的经济性和安全性,因此有必要对燃料组件内定位格架搅混特性进行研究。本文通过粒子图像测速（PIV）技术开展了棒束通道内定位格架上下游流场的可视化研究,对比了有无格架棒束通道内流场的分布特征,定量分析了定位格架对棒束通道流场搅混的贡献。对不同流速下定位格架下游横纵速度的沿程变化特性进行研究,发现了不同流速作用下定位格架对横向、轴向速度的促进和抑制规律。另外,通过速度均方根对下游的湍流特性进行了评估。实验结果可为数值计算提供全场的数据验证,并可为定位格架设计和优化提供基础。     

环形通道内棒偏心和弯曲工况临界热流密度机理研究

本文分别从两种不同类型的临界热流密度（CHF）的触发机理出发,分析了内棒偏心和弯曲对CHF的影响。以氟利昂（R-134a）作为流动工质,在竖直向上流动的环形通道内开展了仅内棒加热的CHF实验研究。实验段包含3种形式：同心、偏心和弯曲。偏心实验结果表明：在高过冷工况下,内棒偏心将对CHF造成惩罚,且偏心率为0783的实验段对CHF惩罚更严重;在低过冷工况下,偏心效应减弱。高压高质量流速工况,空泡漂移效应会导致偏心率为0783的CHF大于偏心率为0435的CHF。弯曲实验结果表明：小闭合度的弯曲对CHF几乎没有影响。大闭合度的弯曲对于低质量流速的Dryout型CHF,弯曲棒会破坏液膜的稳定性;对于低质量流速的DNB型CHF,空泡漂移效应远小于偏心通道,弯曲的CHF小于相同最小间隙下偏心的CHF。     

A non linear model for the PWR fuel assembly seismic analysis

In seismic PWR core analysis, the non linear behavior of fuel assemblies has to be studied taking into account variations of stiffness and damping due to the slippage of rods in spacer grids. Based on the linear CEA’s assembly model, principally composed of two beams representing sets of rods and thimbles, a linear model is described. In the second part of this paper the slippage and loss of contact of rods on a grid relic is analyzed. Based on the Coulomb friction model and plasticity analogy, a grid model constituted of elastoplastic hinge and non linear springs is proposed to simulate the progressive slippage of rods inside the cells. In the third part, a non linear assembly model is designed and validated with experimental data. This model represented with success the increase of damping and decay of frequency when the magnitude of the dynamic loading increases.     

Effects of various spacer grid modeling on the neutronic parameters of the VVER-1000 reactor

The main objective of this paper is to study the effects of various spacer grid models on the neutronic parameters of a VVER-1000 reactor. Specifically, the data of the nuclear power plant at the Bushehr site, which is of a VVER-1000 type, will be studied. Three models, representing the spacer grids along the fuel assemblies are presented. These three models are the homogeneous and the heterogeneous local spacer grid models and the shroud spacer grid model. In the homogeneous and the heterogeneous models, the spacer grids are considered at their actual locations in the axial direction. The only difference between the two models is that in the homogeneous model, the spacer grids are homogenized with the coolant while in the heterogeneous model, the spacer grids are modeled around the fuel cells at their exact axial positions. In the shroud model, the spacer grids are modeled in the shroud region containing the coolant and are not necessarily placed at their appropriate axial positions.     

Vibration analysis of a dummy fuel rod continuously supported by spacer grids

The fuel rods in the pressurized water reactor are continuously supported by a spring system called a spacer grid (SG), which is one of the main structural components for the fuel rod cluster (fuel assembly). The fuel rods have a vibration behavior within the reactor due to coolant flow. Since the vibration, which is called flow-induced vibration, can wear away the surface of the fuel rod, it is important to understand its vibration characteristics. In this paper, a modal testing and a finite element (FE) analysis using ABAQUS on a dummy fuel rod continuously supported by Optimized H Type (OHT) and New Doublet (ND) spacer grids are performed to obtain the vibration characteristics such as natural frequencies and mode shapes and to verify the FE model used. The results from the test and the FE analysis are compared according to modal assurance criteria values. The natural frequency differences between the two methods as well as the mode comparison results for the rod with the OHT SG are better than those with the ND SG. That is, in the case of the ND grid model using beam-spring elements, there was a large discrepancy between the two methods. Thus, we tried to modify the FE model for the ND SG considering the contact phenomena between the fuel rod and the SG. The results of the new model showed a good agreement with the experiment compared with those of a beam-spring model.     

Annular flow experiments in rod bundles with spacers

Measurements of pressure drop, film flow rate and film thickness have been made on a special test section modelling the four subchannels between ten fuel rods in a LWR geometry. Tests were carried out with air/water at 2.5–3-bar (a). Measurements were made with and without spacer grids inserted in the channel. In the former case, the measurements were made upstream and downstream of the spacers. In the case without spacers, the results show similar trends to data from tubular geometries with a similar hydraulic diameter. The two spacers tested both showed similar pressure losses. There was an increase in film flow rate and film thickness downstream of the ULTRAFLOW spacer. For a conventional egg-crate spacer there was a small change, with in some cases a decrease in film flow rate.     

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.     

On the factors affecting the fretting-wear risk of PWR fuel assemblies

Key factors affecting the rod-to-grid fretting-wear risk of fuel assemblies operated in pressurized water reactors (PWR) are evaluated. The analysis is part of a comprehensive approach to predict fretting-wear risk based on the fuel assembly operating conditions. The assembly wear damage is determined by a non-linear vibration model of the nuclear fuel rod exposed to a turbulent flow. The study evaluates the sensitivity of the wear damage to the grid support forces, fuel rod-to-grid gap size, assembly grids misalignment, rod structural damping and stiffness, assembly bow shape, friction coefficients and turbulence force spectrum. The results of the numerical simulations show that the grid cell clearance and the turbulence forces are key factors in the wear process. Since a good correlation exists between these two parameters and the assembly location in the core, it is recommended to include consideration of the wear risk minimization as an additional criterion for the design of the core loading pattern.     

锆合金板织构的控制

针对锆合金条带做ＰＷＲｓ燃料组件定位格架时，元件棒所受的夹持力会因中子辐照引起格架条带伸长而松弛的问题，本文归纳总结了变形温度对锆合金形变机理的影响以及轧制温度和热处理对织构的影响，提出通过改变轧制温度来控制锆合金横向织构取向因子（ＦＴ＝０．４～０．５）是可行的。这样，垂直于轧向取条带制成定位格架后，经中子辐照，格架孔会产生收缩，元件棒夹持力的松弛也可得到补偿。     

Single-phase convective heat transfer enhancement by spacer grids in a rod bundle

The spacer grids within a fuel assembly of a nuclear reactor core disrupt and re-establish the momentum and thermal boundary layers so that they enhance the local heat transfer within and downstream of the spacer grids. An experimental study in a 6×6 rod bundle has been performed to investigate the effects of spacer grids on the single-phase convective heat transfer enhancement. The experimental data showed that the Reynolds number has a significant impact on the heat transfer enhancement only when the Reynolds numbers are lower than about 10,000. The conventional correlations showed poor predictions of the heat transfer enhancement by spacer grids at low Reynolds numbers; in particular, the maximum heat transfer rate at the top end of the spacer grids was significantly overestimated. Furthermore, the conventional correlations did not properly account for the effects of the Reynolds numbers on the heat transfer enhancement. Therefore, more systematic experiments should be performed using various spacer grids with large blockage ratios at low Reynolds numbers, considering an early phase of the reflood conditions.     

Capabilities of TRANSURANUS code in simulating power ramp tests from the IFPE database

Investigations of fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. OECD NEA sets up the “International Fuel Performance Experiments (IFPE) database”, a public domain database on nuclear fuel performance experiments with the purpose of model development and code validation. The objective of the activity (performed in the framework of the IAEA CRP FUMEX-III project) is to investigate the pellet-clad interaction mechanism and the capability of TRANSURANUS code in simulating the phenomena, processes occurring in the fuel rod during the power ramps, with focus on the parameters influencing the cladding failures. The experimental database adopted is the Studsvik PWR Super-Ramp subprogram, part of the IFPE database, which consists of 28 pressurized water reactor fuel rods power ramped at burnup from 28 to 45 MWd/kgU. Relevant results by TRANSURANUS are presented in connection with the experimental evidences. Focus is given on the PCI/SCC failure, demonstrating that the failure threshold, available in TRANSURANUS, results conservative both in case of KWU and W rods.     

定位格架防勾挂与热工性能协同设计研究

定位格架是燃料组件骨架结构的重要组成部分,其主要功能是夹持定位燃料棒,同时还应考虑防勾挂性能和热工性能。本文从主流燃料组件的运行经验反馈出发,利用三维建模软件UG模拟格架相对运动的方法,对产生勾挂的原因进行分析,明确了外条带导向翼采用连续排列能有效提高防勾挂性能,并通过定位格架勾挂试验进行了验证;通过对定位格架所在边栅元内的计算流体力学（CFD）模拟分析,发现基于传统防勾挂设计的导向翼连续排列形式不利于相邻格架之间的热工性能;在此基础上,设计了导向翼高矮交替排列的方案。理论分析和试验验证结果表明,该方案实现了燃料组件定位格架防勾挂与热工性能的协同设计。      

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.     

Analytical study on effects of BWR fuel spacer on droplet deposition

In a boiling water nuclear reactor (BWR), liquid film dryout may occur on a fuel rod surface when the fuel assembly power exceeds the critical power. The spacers supporting fuel rods affect on the thermal-hydraulic performance of the fuel assembly. The spacer is designed to enhance critical power significantly. If spacer effects for two-phase flow could be estimated analytically, the cost and time for the development of the advanced BWR fuel would be certainly decreased. The final goal of this study is to be able to analytically predict the critical power of a new BWR fuel assembly without any thermal-hydraulic tests. Initially, we developed the finite element code to estimate spacer effects on the droplet deposition. Then, using the developed code, the spacer effects were estimated for various spacer geometries in a plane channel and one subchannel of BWR fuel bundle. The estimated results of the spacer effects showed a possibility to analytically predict the critical power of a BWR fuel assembly.     

棒束及格架布置对燃料组件搅混特性影响的实验研究

以CPR1000核电机组使用的格架组装的5×5棒束燃料组件为对象,开展了多组全长棒束燃料组件搅混特性实验,重点分析了冷-热棒布置形式、格架布置形式等几何参数对燃料组件搅混特性的影响规律,实验结果表明,冷-热棒中心对称布置时的燃料组件热扩散系数更接近真值;跨间搅混格架对燃料组件总体热扩散系数有较小增强作用,但对于棒束压降的贡献很低。      

A continuum model for cracked UO2 fuel

This paper presents a constitutive model for uranium dioxide fuel pellets in light water reactor fuel rods. The proposed model accounts for the fuel mechanical behaviour under pellet cracking, fragment relocation and pellet-clad mechanical interaction. Moreover, the detrimental effect of cracking on the fuel thermal conductivity is considered in the model. An essential part of the model is the representation of pellet cracks, which significantly affect both the mechanical and thermal behaviour of nuclear fuel under operation. Cracking is modelled in a continuum context, where cracks are represented by nonelastic strains in the material. The continuum representation is particularly suitable for finite element computer codes, since cracking can be treated in the same manner as plasticity and creep. The model is derived in the form of a nonlinear constitutive relation for the fuel material, that may be implemented in either two- or three-dimensional finite element fuel performance computer codes. The fundamentals of the model are presented, and issues concerning its numerical implementation are discussed. The model's ability to capture important aspects of the cracked fuel behaviour is also illustrated by comparisons with in-reactor experiments.     

The effect of fuel rod loading speed on spacer grid spring force

The burnup-dependent grid-to-rod gap combined with the fluid-induced vibration may generate grid-to-rod fretting wear-induced fuel failure for some fuel assemblies in a certain burnup range. The grid-to-rod gap is dependent on initial spacer grid spring force, spring force relaxation and cladding creepdown. It is found that the initial spring force is reduced during the fuel rod loading into the fuel assembly skeleton. The extent of the initial spring force loss is strongly dependent on the fuel rod loading speed. Based on the initial spring force loss data obtained from two kinds of fuel rod loading speeds of 0.18 and 0.33 m/s, it can be said that the higher rod loading speed generates the larger initial spring force loss. This is because the higher speed generates the larger overshooting of spring deflection during the fuel rod loading. The extent of overshooting may be affected by axial misalignment of SG cells, spring-to-fuel rod end plug contact angle, ballooning of FR end plug weld region and the extent of gravity-induced FR bowing, combining with the fuel rod loading speed. The rod loading speed of 0.33 m/s is found to produce some spacer grid cells less than a minimum initial spring force requirement of 12 N against the grid-to-rod fretting wear-induced failure. In order to produce initial spacer grid spring force meeting the minimum spring force requirement, it is recommended that the lower rod loading speed be used, combined with axially aligned spacer grid cells and lower contact angle of spring-to-fuel rod end plug.