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。     

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.     

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.     

Hydride fuel for LWRs—Project overview

This special issue of Nuclear Engineering and Design consists of a dozen papers that summarize the research accomplished in the DOE NERI Program sponsored project NERI 02-189 entitled “Use of Solid Hydride Fuel for Improved Long-Life LWR Core Designs”. The primary objective of this project was to assess the feasibility of improving the performance of pressurised water reactor (PWR) and boiling water reactor (BWR) cores by using solid hydride fuels instead of the commonly used oxide fuel. The primary measure of performance considered is the cost of electricity (COE). Additional performance measures considered are attainable power density, fuel bundle design simplicity, in particular for BWRs, safety, attainable discharge burnup, and plutonium (Pu) transmutation capability.Collaborating on this project were the University of California at Berkeley Nuclear Engineering Department (UCB), Massachusetts Institute of Technology Nuclear Science and Engineering Department (MIT), and Westinghouse Electric Company Science and Technology Department. Disciplines considered include neutronics, thermal hydraulics, fuel rod vibration and mechanical integrity, and economics.It was found that hydride fuel can safely operate in PWRs and BWRs having comparable or higher power density relative to typical oxide-fueled LWRs. A number of promising applications of hydride fuel in PWRs and BWRs were identified: (1) Recycling Pu in PWRs more effectively than is possible with oxide fuel by virtue of a number of unique features of hydride fuel-reduced inventory of ²³⁸U and increased inventory of hydrogen. As a result, the hydride-fueled core achieves nearly double the average discharge burnup and the fraction of the loaded Pu it fissions in one pass is double that of the MOX fuel. (2) Eliminating dedicated water moderator volumes in BWR cores, thus enabling significant increase of the cooled fuel rod surface area as well as the coolant flow cross-section area in a given fuel bundle volume while reducing the heterogeneity of BWR fuel bundles, thus achieving flatter pin-by-pin power distribution. The net result is an increase in the core power density and a reduction of the COE.A number of promising oxide-fueled PWR core designs were also found in this study: (1) The optimal oxide-fueled PWR core design features a smaller fuel rod diameter (D) of 6.5 mm and a larger pitch to rod diameter (P/D) ratio of 1.39 than that presently practiced by industry of 9.5 mm and 1.326. This optimal design can provide a 27% increase in the power density and a 19% reduction in the COE provided the PWR can be designed to have the coolant pressure drop across the core increased from the reference 0.20 MPa (29 psi) to 0.414 MPa (60 psi). Under the set of constraints assumed in this work, hydride fuel was found to offer comparable power density and economics as oxide fuel in PWR cores when using fuel assembly designs featuring square lattice and grid spacers. This is because pressure drop constraints prevented achieving sufficiently high power using hydride fuel with a relatively small P/D ratio of around 1.2 or less, where it offers the highest reactivity and a higher heavy metal (HM) loading. (2) Using wire-wrapped oxide fuel rods in hexagonal fuel assemblies, it is possible to design PWR cores to operate at ∼50% higher power density than the reference PWR design that uses grid spacers and a square lattice, provided 0.414 MPa coolant pressure drop across the core could be accommodated. Uprating existing PWRs to use such cores could result in up to 40% reduction in the COE. The optimal lattice geometry is D = 9.34 mm and P/D = 1.37. The most notable advantages of wire-wraps over grid spacers are their significantly lower pressure drop, higher critical heat flux, and improved vibration characteristics.The achievement of the highest power gains claimed in this study is possible as long as mechanical components like assembly hold-down devices (both in PWRs and in BWRs) and steam dryers (only in BWRs) are appropriately upgraded to accommodate the higher coolant pressure drop and flow velocities required for the high-performance LWR designs. The compatibility of hydride fuel with Zircaloy clad and with PWR and BWR coolants need yet be experimentally demonstrated. Additional recommendations are given for future studies that need to be undertaken before the commercial benefits from use of hydride fuel could be reliably quantified.     

燃料组件5×5格架多跨模型CFD模拟方法研究

本文详细描述了某典型燃料组件5×5格架模型CFD分析的几何模型简化、网格划分、求解及后处理等过程。在5×5结构单跨模型上研究了弹簧刚突对搅混特性及压降的影响,并采用简化弹簧刚突的5×5格架模型实现了包含11层格架的多跨模型计算。单跨模型计算结果表明,弹簧刚突结构强化了横向流动,利于换热,Nu提高了8%,但弹簧刚突格架模型较简化弹簧刚突模型压降损失增加了40%。多跨模型计算得到了多层格架全程流动换热特性,为燃料组件自主研发中定位格架数量及布置的设计优化以及DNB预测计算提供了有效的CFD分析方法。     

Void-fraction distribution under high-pressure boil-off conditions in rod bundle geometry

Void fractions in a simulated pressurized water reactor (PWR) core rod bundle geometry were measured under boil-off conditions covering pressures from 3 to 12 MPa and mass fluxes from 5 to 100 kg m⁻² s⁻¹, with a particular interest in void fractions at higher pressures and relatively high mass fluxes. Test results showed that the Chexal-Lellouche model predicts best the present (volume-averaged) void-fraction data among correlations and models examined in this study. The volume-averaged void fractions obtained from differential pressure measurements are systematically smaller than the chordally averaged void fractions obtained from gamma densitometer measurements. Local void fractions were measured in the same bundle for non-heated steam-water two-phase flow of 3 MPa by using an optical void probe. It was found that the difference between the volume-averaged and chordally averaged void fractions mentioned above can be explained qualitatively by a local void-fraction distribution in the bundle measured in the latter tests.     

Preliminary study of parameter uncertainty influence on Pressurized Water Reactor core design

Sub-channel analysis can improve the accuracy of reactor core thermal design. However, the important initial parameters contain various uncertainties during reactor operation. In this work, the Sub-channel Analysis Code of Supercritical reactor (SACOS) code, which is also applicable for Pressurized Water Reactor (PWR), was used to study the coolant flow characteristic and fuel rod heat transfer characteristic of 1/8 assembly which has the maximum linear power density in 300 MWe PWR core firstly. Then the Wilks' method and Response Surface Method (RSM) were utilized to determine the influence of sub-channel input parameters uncertainties on the highest temperature of reactor core fuel rod and Minimum Departure from Nucleate Boiling Ratio (MDNBR). The results show that in the most conservative conditions, the maximum temperature of the fuel rod and MDNBR were 2167.4 °C and 1.08, respectively. Considering the uncertainties of assembly inlet flow rate, inlet coolant temperature and system pressure, the 95% probability values (with 95% confidence) of fuel rod maximum and MDNBR calculated using response surface methodology were 2144.0 °C and 1.6, while they were 2137 °C and 1.74 calculated by Wilks' approach. Results show that the uncertainty analysis methods can provide larger reactor design criteria margin to improve the economy of reactor. Furthermore, the code was developed to have the capacity to perform the uncertainty study of sub-channel calculation.     

Prospects for fission gas extraction from in-reactor UO2 nuclear fuel elements

Due to the many problems encountered in the design of fuel rods for the safe operation of commercial nuclear reactors, caused by the fission gases generated by the fission of fissile material, it was considered opportune to make a theoretical analysis of the feasibility of extraction of fission gases from the fuel rod while in operation.This analysis in the steady state of a Zircaloy-2 sheathed fuel rod containing UO₂ as a fuel, with a 2 mm (2.7 vol.%) diameter porous graphite cylinder inserted in the centre, has demonstrated that a total volume of fission gases (xenon, krypton, and iodine) of about 1.1 × 10⁻⁶ cm³/s (at STP) can be extracted from the fuel rod at a controlled rate, determined by the inherent property of fission gas migration towards the centre of the fuel rod from its place of formation. In this analysis, the fuel rod was assumed to be subjected to irradiation in a reactor the size of a Bruce “A” reactor, operating at 3000 megawatts thermal power. The extracted volume of gas was calculated on a 900 h cycle after the first 90 h of reactor operation had elapsed.     

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。     

超临界水堆燃料棒流致振动简化模型

在超临界水堆中,当超临界水流过带有绕丝的燃料棒时可能诱发其发生振动,使得燃料包壳发生疲劳现象。带有的接触的非线性有限元模型使得计算量大大增加,而且其计算精度仍有待实验验证。本文针对超临界水堆流致振动实验,将绕丝的影响简化为弹簧,建立燃料棒流致振动的简化模型,并通过有限元模型对燃料棒的固有特性进行分析,验证了模型的正确性。最后,以功率谱对模型加载,求得了超临界水堆燃料棒的位移响应和1δ解。     

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。     

CFD analyses in tight-lattice subchannels and seven-rod bundle geometries of a Super Fast Reactor

This paper presents CFD analyses in heat unsymmetric subchannels and heat symmetric seven-rod bundle geometries of a Super Fast Reactor (Super FR) fuel assembly using STAR-CD. The purpose of CFD analyses in heat unsymmetric subchannels is to evaluate the effect of the power differences on the heat transfer in subchannels of the Super Fast Reactor. For heat symmetric seven-rod bundles, the effects of the gap clearance between the fuel rod and the assembly wall and the displacement of the fuel rod on the circumferential temperature distributions and Maximum Cladding Surface Temperature (MCST) are analyzed. The results show that larger power difference between fuel rods gives larger circumferential temperature difference of the hottest fuel rods. Considering cross flow between edge and ordinary subchannels, 1 mm gap between the fuel rod and the assembly wall is better for small MCST although the circumferential temperature difference in edge subchannel is large. MCST increases exponentially with the displacement. The relative error of displacement should be less than 1% if the allowable increment of MCST due to displacement is less than 6 °C.     

A power perturbation technique for the measurement of fuel-to-coolant heat transfer coefficients in fast reactors

This technique provides a method of obtaining average fuel to coolant heat transfer coefficients for individual fuel subassemblies in fast reactors. A series of experiments on the UK prototype fast reactor (PFR) over the period 1977–1979 have demonstrated that the technique is simple, requires no special instrumentation other than thermocouples to monitor coolant outlet temperatures, and the measurement can be made during normal reactor operation. Thus it is possible to determine how heat transfer coefficients change with operating conditions and with the degree of burn-up in the fuel.The analysis of a single experiment is presented to illustrate the technique. This was conducted at a single reduced power level of 200 thermal megawatts for two different primary coolant flow rates, both steady fractions of the maximum (0.88 and 0.47). Cyclic and single-step perturbations of about 10% amplitude were impressed on the steady power and the delayed coolant temperature response at subassembly outlets was monitored. Burn-ups in the subassemblies ranged between 1.0% and 4.7%. From the measured delays at the two flows it was possible to determine the fuel time-constant and hence the fuel-to-coolant heat transfer coefficient. It was also shown that a simple, lumped-element, heat transfer model can be used to obtain sufficiently accurate estimates from measurements at just one coolant flow.Fuel surface-to-coolant thermal conductances (i.e. gap conductances) were subsequently derived from the heat transfer coefficients. These ranged between 2.4 kW m⁻² K⁻¹ and 3.3 kW m⁻² K⁻¹ with the smaller conductances being obtained for those fuel elements with the larger degree of burn-up. These values are lower than expected but consistent with a higher than expected value for the negative power coefficient of reactivity feedback which has been observed at reduced power.     

The effect of fuel rod supporting conditions on fuel rod vibration characteristics and grid-to-rod fretting wear

The grid-to-rod fretting wear-induced fuel rod failure observed in PWRs may be caused by excessive fluid-induced vibration and inadequate fuel rod support by the spacer grid spring. In order to simulate in-reactor grid-to-rod fretting wear behaviors, the grid-to-rod fuel rod supporting conditions as a function of time were predicted by taking into account cladding creep rate, initial spacer grid spring deflection, spacer grid spring force relaxation, etc. Based on these grid-to-rod supporting conditions, the fuel rod vibration modes and natural frequencies were calculated with the help of the ANSYS code, while the fuel rod vibration amplitudes were estimated by the Paidoussis’ empirical formula. With these vibration characteristics that depend upon the grid-to-rod supporting conditions, the in-reactor fretting wear axial profile observed on the fuel rod surface are found to be simulated quite well. In addition, key design guidelines for the fuel assembly and the spacer grid are proposed to minimize the grid-to-rod fretting wear that may be utilized to develop an advanced fuel design against fretting wear.     

Experimental research progress on critical heat flux of Chinese PWR

One of the most important requirements in the design of pressurized water reactor (PWR) is to avoid the occurrence of critical heat flux (CHF). The design criteria for PWR specify that they must be operated at a certain percentage below CHF at all times and locations so as to the cladding temperature of fuel element at safe values. So in the process of safety assessment, CHF is one of important thermal-hydraulic parameters limiting the available power, whose size directly affects safety and economy of PWR nuclear power plant. This paper deals with a summary of experimental research progress on CHF of Chinese PWR. It mainly presents CHF experimental researches of Φ10 fuel assembly, CHF experimental researches of standard fuel assembly, and CHF experimental progress of non-uniform heated rod bundles. It should be emphasized that it also presents experimental research programs on CHF of Chinese advanced fuel assembly with self-reliance copyright. All CHF data obtained will be used for design improvement of Chinese PWR and R&D program of New Generation 1000 MWe PWR.     

A method of predicting the temperature response of ballooning fuel rod cladding for PWR LOCA conditions

A method is described for calculating fuel rod cladding temperatures in a blockage formed by a group of ballooned fuel rods in a larger rod array, for heat transfer conditions appropriate to the reflooding phase of a postulated PWR LOCA. The model is suitable for describing the extreme case of co-planar axially extended balloons, where steam superheating and skin friction effects are believed to have an important effect on blockage heat removal. Attention is restricted to the constricted zone within the blockage.Reasonable agreement is shown with available heat transfer data from partially ballooned rod arrays, for conditions of steam cooling, steam-and-droplet cooling and reflood cooling. The model is also able to describe flow velocity distribution data from partially blocked rod bundles with reasonable accuracy.Parametric calculations for typical PWR LOCA heat transfer conditions suggest that blockage length has a strong effect on fuel coolability, mainly as a result of extra superheating of the steam within the blockage. However calculations also indicate that the presence of entrained water droplets has a powerful effect in reducing the clad temperatures attained.     

Reflooding experiments with LWR-type fuel rod simulators in the QUENCH facility

The QUENCH experiments at the Karlsruhe Research Center are carried out to investigate the hydrogen generated during reflooding of an uncovered Light Water Reactor (LWR) core. The QUENCH test bundle is made up of 21 fuel rod simulators approximately 2.5 m long. The Zircaloy-4 rod cladding is identical to that used in PWRs (Pressurized Water Reactors) with respect to material and dimensions. Pellets are made of zirconia to simulate UO₂. After a transient phase with a heating rate of 0.5–1 K s⁻¹ water of approx. 395 K is admitted from the bottom when the test bundle has reached its pre-determined temperature. Except for the flooding (quenching) phase, the QUENCH test phases are conducted in an argon/steam atmosphere at 3 g s⁻¹ each. The results of the first two experiments, QUENCH-01 (with pre-oxidation of 300 μm oxide layer thickness at the cladding outside surface) and QUENCH-02 (reference test without pre-oxidation), are compared in the paper. The pre-oxidized LWR fuel rod simulators of QUENCH-01 were quenched from a maximum temperature of 1870 K. In the second bundle experiment, QUENCH-02, quenching started at 2500 K. Pre-oxidation apparently prevented a temperature escalation in the QUENCH-01 test bundle, while the QUENCH-02 test bundle experienced a temperature excursion which started in the transient phase and lasted throughout the flooding phase. The different behavior of the two experiments is also reflected in hydrogen generation. While the bulk of H₂ was produced during pre-oxidation of test QUENCH-01 (30 g), the largest amount, i.e. 170 g, of hydrogen was generated during the reflooding phase of test QUENCH-02, at a maximum production rate of 2.5 g s⁻¹ as compared to 0.08 g s⁻¹ in test QUENCH-01. Similarities between the two experiments exist in the thermo-hydraulics during the quench phase, e.g. in the cooling behavior, the quench temperatures, and quench velocities.     

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.     

5×5花瓣形燃料棒束组件内单相流动与换热特性数值模拟研究

相较于传统圆柱形燃料棒,花瓣形燃料棒具有安全裕量高等优点,研究其在压水堆运行工况下的热工水力特性具有重要意义。本文通过STAR-CCM+对5×5花瓣形燃料棒束组件进行数值模拟研究,计算并分析了组件内二次流速度、温度、换热系数等关键热工参数,获得了入口流速、螺旋节距对组件内部流动与换热特性的影响规律。计算结果表明：花瓣形燃料棒的螺旋结构可增强冷却剂的横向流动,同一高度上燃料棒表面温度分布具有周期性,增大入口流速可增强燃料棒的表面换热,消除温度分布的不均匀性。此外,螺旋节距大于750 mm,燃料棒换热性能与无扭转的燃料棒相差不大,甚至更低。     

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。