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.     

The study on grid-to-rod fretting wear models for PWR fuel

The fretting wear is found to be generated at grid-to-rod contact areas by flow-induced vibration. This flow-induced grid-to-rod fretting wear may be initiated at a certain critical grid-to-rod gap that strongly depends on the extent of flow-induced vibration and grid spring designs. Three fretting wear excitation mechanisms acting on the grid-to-rod fretting wear are summarized. In order to examine the impact of grid spring designs on the fretting wear rate, the fretting wear tests for three kinds of grid spring designs were carried out for 500 h, simulating the reactor flow conditions. In parallel, three grid-to-rod fretting wear models that include constant work rate model, constant work density rate model and linear work density rate model have been developed. The three fretting wear models were used to predict the fuel rod perforation times with the use of the fretting wear test results. It is said that the constant work density rate model or the linear work density rate model is quite effective in predicting the grid-to-rod fretting-induced rod failure time observed in commercial nuclear power plants.     

A study on the grid-to-rod fretting wear-induced fuel failure observed in the 16×16KOFA fuel

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 systematic grid-to-rod fretting wear-induced fuel failure occurred at the 16×16 Korean Optimized Fuel Assembly loaded in the 2-loop Westinghouse type plant in Korea. Prior to various tests and some measurements for investigating its root causes, they were assumed to be self-excited fuel assembly vibration caused by hydraulic-unbalanced mixing vane design, excessive cross-flow between fuel assemblies during the transition core, or relatively large grid-to-rod gap formation during in-reactor irradiation that may be caused by excessive initial spring force loss of fresh fuel during a fuel rod loading process and/or a fuel assembly transport to a plant and by excessive cladding creep-down. A wide spectrum of tests and some measurements were performed to find out root cause(s) of the grid-to-rod fretting wear-induced fuel failure. Based on these tests and measurements, it is concluded that the self-excited fuel assembly vibration is the primary root cause, while excessive initial spring force loss during the fuel rod loading process is the second major root cause.     

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.     

Optimization of a nuclear fuel spacer grid spring using homology constraints

The springs in a spacer grid support the fuel rods in a nuclear fuel system. The spacer grid is part of the fuel assembly. Since the spring has repeated contact with the fuel rod, fretting wear occurs on the surface of the fuel rod. Design is usually performed to reduce the wear while the functions of the spring are maintained. The design process for the spring is defined by using the Independence Axiom of axiomatic design and the design is carried out based on the direction that the design matrix indicates. For a detailed design, an optimization problem is formulated. In optimization, a homologous design is employed to reduce the fretting wear. The deformation of a structure is called homologous if a given geometrical relationship holds for a certain number of structural points before, during, and after the deformation. In this case, the deformed shape of the spring should be the same as that of the fuel rod. This condition is transformed to a function and considered as a constraint in the optimization process. The fretting wear is expected to be reduced due to the homology constraint. The objective function is minimizing the maximum stress to allow local plastic deformation. Optimization results show that contact occurs in a wide range. The results are verified by non-linear finite element analysis.     

On the geometry of the fuel rod supports concerning a fretting wear failure

Geometrical conditions of spacer grid springs and dimples of a light water reactor fuel assembly are studied in this paper concerning a fuel rod’s fretting wear failure. In this framework, the springs/dimples are categorized from the aspects of their orientation with respect to the fuel axis and the contact types. Possible motions on the contacts between the springs/dimples and fuel rods are estimated by conducting a flow-induced vibration test. Features of the wear scar and depth are investigated by independent fretting wear tests carried out with spring and dimple specimens of typical contact geometries. It is also attempted here to apply the contact mechanics theory to a fuel fretting wear analysis such as the prediction of a wear depth profile and its rate, which is influenced by the contact shape of the springs/dimples. It is shown that the theory can be applied to a dimensional control of a coining for the springs/dimples, which is usually carried out in a thin plate fabrication. From the results, the necessary conditions for a spring/dimple geometry for restraining a fretting wear failure are discussed.     

Development of an advanced PWR fuel for OPR1000s in Korea

The advanced PWR fuel for the OPR1000s in Korea, PLUS7, has been developed to enhance thermal performance, high burnup capability and fuel reliability against grid-to-rod fretting wear and debris. The outstanding design features of PLUS7 include mixing vane mid-grids for increasing thermal performance and minimizing vibration-induced fretting wear, optimized fuel dimensions and advanced zirconium alloys for high burnup capability of 72,000 MWD/MTU, and an optimized fuel rod diameter for reducing pressure drop and improving neutron economy. The fuel assembly and its components performances have been verified through a wide spectrum of mechanical, thermal hydraulic, vibration and fretting wear tests. Based on the verification test results and the evaluations with the help of the KNF design code system, it is found that the PLUS7 fuel will maintain its integrity up to the envisaged burnup of 72,000 MWD/MTU. In addition, the PLUS7 fuel performances were evaluated to be considerably improved in comparison with the current fuel used in the OPR1000s.     

Mechanical analysis of fuel fretting problem

Fuel fretting is studied by contact mechanics approach. Shear force produced by flow-induced vibration is regarded as the major factor of the fuel fretting. Contact dimension is examined for the Korean PWR Fuel Assembly using finite element method. Axial direction is incorporated with transverse one for the shear force. As for the sequence of the shear, a closed rectangular as well as an oblique path are considered to simulate the actual behaviour due to the vibration. The shear stresses on the contact surface between fuel rod and spacer grid is evaluated numerically. It is supposed that a partial slip regime prevails on the contact at the early stage of fuel life. In case of gross slip, the present method can be applied without modification. The dissipation of friction energy on the contact is evaluated and discussed for a wear model and a spacer grid 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。     

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.     

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。      

Axial-flow-induced vibration for a rod supported by translational springs at both ends

An axial-flow-induced vibration model was proposed for a rod supported by two translational springs at both ends in order to evaluate the sensitivity to spring stiffness on the FIV for a PWR fuel rod. For developing the model, a one-mode approximation was made based on the assumption that the first mode was dominant in vibration behavior of the single span rod. The first natural frequency and mode shape functions for the flow-induced vibration, called the FIV, model were derived by using Lagrange's method. The vibration displacements were calculated by both of the spring-supported rod and the simple-supported (SS) one. As a result, the vibration displacement for the spring-supported (50 kN m⁻¹) rod was 15–20% larger than that of the SS rod when the rods are in axial flow of 5–8 m s⁻¹ velocity. The discrepancy between both displacements became much larger as flow velocity increased, and that of the rod having the short span length was larger than that of the rod having the long span length although the displacement value itself of the long span rod was larger than that of the short one. The vibration displacement for the spring-supported rod appeared to decrease with the increase of the spring constant. Since single span beam supported by the two translational springs are focused on in this paper, further study will be needed to reflect more realistic supporting conditions of the PWR fuel rod such as two springs and four dimples and cross or swirling flow caused by the mixing vane of the spacer grid.     

FOCUS-type fuel assembly for PWRs

In recent years the development efforts for Siemens PWR fuel assemblies were mainly concentrated on reducing the fuel cycle costs and increasing the operational reliability of the fuel assemblies.The first objective was aimed at increasing the average discharge burnup to > 50 MWd/kgU and increasing the critical heat flux. The high envisaged burnup required to develop a corrosion resistant cladding tube outside the Zry-4 range. The decision was made to use a Duplex cladding tube consisting of a corrosion optimized outer layer on a Zry-4 base material. A ZrSnFeCr alloy with reduced tin content was chosen for the outer layer. The critical heat flux could be increased by introducing mixing vanes on the spacer grids within the active length.To reach the second objective, reliable avoidance of spacer grid damage during core loading and unloading and reduction of fuel rod defects by debris fretting, the spacer grid corners were improved and a debris separation grid was developed.These design improvements were introduced into the new FOCUS-type fuel assembly. The name FOCUS stands for “Fuel assembly with Optimized Cladding and Upgraded Structure”.     

Impact of thicker cladding on the nuclear parameters of the NPP Krško fuel

To make fuel rods more resistant to grid-to-rod fretting or other cladding penetration failures, the cladding thickness could be increased or strengthened. Implementation of thicker fuel rod cladding was evaluated for the NPP Krško that uses 16 × 16 fuel design. Cladding thickness of the Westinghouse standard fuel design (STD) and optimized fuel design (OFA) is increased. The reactivity effect during the fuel burnup is determined. To obtain a complete realistic view of the fuel behaviour a typical, near equilibrium, 18-month fuel cycle is investigated. The most important nuclear core parameters such as critical boron concentrations, isothermal temperature coefficient and rod worth are determined and compared.     

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.     

Indirect estimation method of the turbulence induced fluid force spectrum acting on a fuel rod

This paper proposes a methodology to identify a turbulent flow induced force acting on a nuclear fuel rod based on the indirect input force estimation theory in structural dynamics, which is useful to predict the forcing function when the input force cannot be measured directly. Since the nuclear fuel rod in a PWR (pressurized water reactor) is exposed to coolant flow, the turbulence induced force generates a fuel rod vibration which may cause a fretting wear on the surface of the rod. This study develops a method to estimate turbulence induced force spectrum indirectly for a real scale fuel rod loaded in a nuclear fuel test facility. The proposed method requires a reliable finite element (FE) model which simulates the fuel rod dynamics well; therefore, the FE model is discussed, especially regarding the procedure to determine the effective rod density. Since the pellets rattle inside the tube due to small gaps between the tube and pellets, especially at the beginning of the fuel's life, the contribution of the pellet mass to the density for the FE model cannot be determined clearly. It is shown that the appropriate density can be estimated by comparing the natural frequencies from the modal test results of the rod (with pellet) and the tube (without pellet). Then, the indirect turbulence induced force estimation theory is applied to the fuel rod, and some numerical and test results are discussed to verify the applicability of the suggested method.     

A study on mechanical properties and flow-induced vibrations of coil-shaped holddown spring

The fuel assemblies used in the OPR1000s in Korea employ four coil-shaped hold-down springs to exert compressive load at the top of fuel assembly so that the assemblies may not be damaged by preventing its hydraulic-induced lifting-off from its lower seating surface. However, the coolant flow generates the flow-induced vibration at the coil-shaped hold-down springs which may cause wear on the spring surfaces. A hold-own spring may be fractured if torsional stress acting on its worn area exceeds a stress limit, resulting in the loss of hold-down spring force of the fuel assembly. In this paper, flow-induced vibration tests were performed for standard and improved coil type hold-down springs to investigate the effects of these two hold-down spring designs on flow-induced vibration wear. In parallel, a wide spectrum of mechanical tests was performed to obtain vibration-related characteristics of these two hold-down springs, which can be used as input data for the fuel assembly static and dynamic analysis. It is found that the improved hold-down spring design is better against flow-induced vibration wear than the standard one. With the use of the three-dimensional Solidwork model, the stress-related design lifetime of the improved hold-down spring was estimated by extrapolating its wear data measured from the flow-induced vibration tests, which indicates that the improved HD spring design will maintain integrity during the fuel design lifetime in OPR1000s in Korea.     

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.     

格架松弛对燃料棒湍流激励及微振磨损的影响研究

假设所有支承有效,基于燃料棒模态分析的结果,根据压水堆燃料棒的流场分布特征,采用功率谱密度表征湍流激励,结合相关功率谱密度试验参数,求解了各阶模态的振动位移均方值,基于ARCHARD磨损公式计算了燃料棒刚凸位置的磨损深度。由于制造工艺、运输、辐照的影响,格架对燃料棒的夹持作用可能松弛。依次假设格架单个刚凸及弹簧松弛,研究了松弛对燃料棒模态、流致振动以及磨损的影响。结果表明：格架弹簧的松弛对固有频率的影响可忽略;原振幅较大的位置附近刚凸松弛对固有频率影响明显;堆芯入口及出口的横向流速较大,燃料棒底部和顶部的湍流激励振幅较大,这些位置的刚凸支承松弛使湍流激励振幅明显增大,中间位置的刚凸支承松弛对振幅影响较小;刚凸支承松弛对磨损深度的影响与对湍流激励最大振幅的影响趋势基本一致。磨损除了与湍流激励振幅相关,还与固有频率相关,顶部振型和频率乘积的影响大于底部格架位置,顶部格架刚凸松弛对磨损影响最大。     

Mechanical Performance of Newly Developed Supporting Structure of Spacer Grid

The fuel rod supporting structure of spacer grid is of great importance to ensure the fuel rod at the appropriate horizontal and axial position. On the purpose of improving the neutronic economy and simplifying the manufacturing and assembling process of spacer grid, the innovative curved dimple and vaulted spring with small dimple on it were designed, which were directly punched from the zirconium strap of spacer grid. The mechanical experiments were performed to acquire the deformation-load curves of dimples and springs of single strap and in-grid cells, which could provide the load and residual deformation at 100% and 120% nominal designed deformations. The result demonstrates that the designed supporting structure has a relative stable load at the 100% nominal designed deformation, and the test method of the single strap is representative for the similar deformation-load curve between the spring and the in-grid cell.