Experimental and numerical investigations of three-dimensional turbulent flow of water surrounding a CANDU simulation fuel bundle structure inside a channel

Computational fluid dynamics (CFD) is used to simulate highly turbulent coolant flows surrounding a simulation CANDU^® fuel bundle structure inside a flow channel. Three CFD methods are used: large eddy simulation (LES), detached eddy simulation (DES), and Reynolds stress model (RSM). The outcome of the simulations is compared with the experimental pressure data measured using an in-water microphone and a miniature pressure transducer placed at various locations in the vicinity of the bundle structure. Among all the three methods employed in developing computational models, LES provides the most accurate results for turbulent pressures.     

Assessment of core characteristics during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

A transition from 37-element natural uranium fuel to CANFLEX-NU fuel has been modeled in a 1200-day time-dependent fuel management simulation for a CANDU 6 reactor. The simulation was divided into three parts. The pre-transition period extended from 0 to 300 FPD, in which the reactor was fuelled only with standard 37-element fuel bundles. In the transition period, refueling took place only with the CANFLEX-NU fuel bundle. The transition stage lasted from 300 to 920 FPD, at which point all of the 37-element fuel in the core had been replaced by CANFLEX-NU fuel bundle. In the post-transition phase, refueling continued with CANFLEX-NU fuel until 1200 FPD, to arrive at estimate of the equilibrium core characteristics with CANFLEX-NU fuel. Simulation results show that the CANFLEX-NU fuel bundle has a operational compatibility with the CANDU 6 reactor during the transition core, and also show that the transition core from 37-element natural uranium fuel to CANFLEX-NU can be operated without violating any license limit of the CANDU 6 reactor.     

Static strength analysis of CANDU-6 reactor fuel bundle

A static analysis, finite-element (FE) model was developed to simulate out-reactor fuel–string strength tests with use of the well-known, structural analysis computer code ABAQUS. The FE model takes into account the deflection of fuel elements, and stress and displacement in endplates subjected to hydraulic drag loads. It was adapted to the strength tests performed for CANFLEX 43-element bundles and the existing 37-element bundles. The FE model was found to be in good agreement with experiment results. With use of the FE model, the static behavior of the fuel bundle string, such as load transfer between ring elements, endplate rib effects, hydraulic drag load incurring plastic deformation in fuel string and hydraulic flow rate effects were investigated.     

Benchmark Simulation of Turbulent Flow through a Staggered Tube Bundle to Support CFD as a Reactor Design Tool. Part II: URANS CFD Simulation

In Part II, we described the unsteady flow simulation and proposed a modification of a traditional turbulence flow model. Computational fluid dynamics (CFD) simulations of an isothermal, fully periodic flow across a tube bundle using unsteady Reynolds averaged Navier-Stokes (URANS) equations, with turbulence models such as the Reynolds stress model (RSM) were investigated at a Reynolds number of 1.8 × 10⁴, based on the tube diameter and inlet velocity. As noted in Part I, CFD simulation and experimental results were compared at five positions along (x; y) coordinates. The steady RANS simulation showed that four diverse turbulence models were efficient for predicting the Reynolds stresses, and generally, SRANS results were marginal to poor, using a consistent evaluation terminology. In the URANS simulation, we modeled the turbulent flow field in a manner similar to the approach used for large eddy simulation (LES). The time-dependent URANS results showed that the simulation reproduces the dynamic stability as characterized by transverse oscillatory flow structures in the near-wake region. In particular, the inclusion of terms accounting for the time scales associated with the production range and dissipation rate of turbulence generates unsteady statistics of the mean and fluctuation flow. In spite of this, the model implemented produces better agreement with a benchmark data set and is thus recommended.     

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。     

Estimation of fuel and clad temperature of a research reactor during dry period of de-fuelling operation

Heat transfer and fluid flow studies related to spent fuel bundle of a research reactor in fuelling machine has been carried out. When the fuel is in reactor core, the heat generated in the fuel bundle is removed by heavy water under normal reactor operation. However, during the de-fuelling operation, the fuel bundle is exposed to air for some period called dry period. During this period, the decay heat from fuel bundle has to be removed by air flow. This flow of air is induced by natural convection only. In this period, the temperatures of fuel and clad rise. If clad temperature rises beyond a certain limit, structural failure may occur. This failure can result into release of fission products from fuel rod. Hence the temperature of clad has to be within specified limit under all conditions. The objective of this study is to estimate the clad temperature rise during the dry period.In the CFD simulation, the turbulent natural convection flow over fuel and radiation heat transfer are accounted. Standard k-? model for turbulence, Boussinesq approximation for computing the natural convection flow and IMMERSOL model for radiation are used.The steady state and transient CFD simulation of flow and heat is performed, using the CFD code PHOENICS. The steady state analysis provides the maximum temperature the clad will attain if fuel bundle is left exposed to air for sufficiently long time. For safe operation, the clad temperature should be limited to a specified value. From steady state CFD analysis, it is found that steady state clad temperature for various decay powers is higher than the limiting value. Hence transient analysis is also performed. In the transient analysis, the variation of clad temperature with time is predicted for various decay powers. Safe dry time, i.e. the time required for clad to reach the limiting value, is predicted for various decay powers. Determination of safe dry time helps in deciding the time available to the operator to drop the bundle in light water pool for storage. The analysis is found useful in optimizing the de-fuelling process.     

蒸汽发生器传热管束过冷沸腾区两相流动数值模拟

基于两流体欧拉数学模型结合RPI壁面沸腾模型,利用大型商用CFD软件ANSYS CFX 12.0对蒸汽发生器传热管束过冷沸腾区一次侧、壁面和二次侧耦合传热过程进行了数值模拟。研究了三叶梅花孔支撑板和不同入口过冷度条件下蒸汽发生器传热管束内的流动沸腾现象,得到一、二次侧流场与温度场,二次侧空泡份额分布,支撑板梅花孔局部的流动状况及不同入口过冷度对蒸汽发生器热工水力特性的影响。数值模拟结果表明,三叶梅花孔支撑板的存在及不同入口过冷度对蒸汽发生器传热管束过冷沸腾区域的热工水力特性影响显著。     

熔盐球床堆堆芯入口热工水力特性数值分析

堆芯入口流场设计是小型固态燃料熔盐堆系统项目内容之一,它对反应堆结构的稳定性、堆芯温度和流场分布有着非常重要的影响。研究了熔盐流道流通面积变化对堆芯入口温度、流场分布及压降的影响,优化熔盐流道几何结构。以小型熔盐球床堆模型为研究对象,取符合实际边界条件的输入参数,通过改变熔盐流道流通面积,使用计算流体力学(Computational Fluid Dynamics,CFD)通用程序Fluent 16.0对堆芯入口内熔盐的热工水力特性进行数值模拟。在考虑实际下反射层流道的流通面积占比最大为18.14%下,研究了熔盐流道流通面积占比在区间[0,15.00%]变化。结果表明,堆芯活性区熔盐最高局部热点温度随熔盐流道流通面积比的增大而增高;堆芯入口内的压降随下反射层熔盐流道流通面积比的减小而增大;在径向方向上流进孔道的熔盐流速随着孔道远离堆芯位置而增大。本研究可为小型固态燃料球床熔盐堆优化设计提供一定的参考价值。     

含弯曲棒的全长燃料棒束沸腾传热CFD计算分析

采用计算流体动力学（CFD）分析方法模拟了含一根弯曲燃料棒（简称“弯曲棒”）的5×5全长燃料棒束内的沸腾传热现象。基于欧拉两流体模型和改进的壁面沸腾模型进行计算,并基于压水堆子通道和棒束实验( PSBT )基准题中的试验数据对计算方法进行了验证,计算所得截面平均空泡份额与试验数据吻合良好,说明了现有计算方法的可靠性。基于计算结果考察了弯曲棒对棒束通道内流场、温度场、空泡份额等关键参数的影响。研究结果表明,弯曲棒的存在对截面横向流动、流体温度、空泡份额等均未产生显著影响,但弯曲棒表面温度增加,气泡也易发生聚集,增加了发生临界热流密度（CHF）的风险。      

Evaluation of a numeric procedure for flow simulation of a 5 × 5 PWR rod bundle with a mixing vane spacer

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.     

使用离散钍铀燃料组件的CANDU6堆物理特性初步研究

为研究钍铀燃料在CANDU6堆中的应用,采用DRAGON/DONJON程序,对使用离散型钍铀燃料37棒束组件的CANDU6堆进行时均堆芯分析。结果表明,组件采用235U富集度为2.5%的铀棒以及第1、2、3圈布置钍棒的37棒束组件,堆芯在8棒束换料、3个燃耗分区的方案下,组件的冷却剂空泡反应性较使用天然铀的37棒束组件（NU-37组件）与采用混合钍铀元件棒的37棒束组件更负;堆芯最大时均通道/棒束功率满足小于6700?kW/860?kW的限值;燃料转化能力比采用NU-37组件时更高;卸料燃耗可到达13400?MW·d/t(U)。研究表明,所设计的离散型钍铀燃料37棒束组件可用于现有CANDU6堆芯,且无需对堆芯结构及控制机构作重大改造;燃料组件和堆芯设计方案可为钍铀燃料在CANDU6堆芯的应用提供参考。      

Void Fraction in Simulated PWR Fuel Bundle during Reflood Phase

In order to evaluate void fraction in a bundle geometry during the reflood phase, reflooding experiment with a 4×4 simulated fuel array was conducted.

As the result, it was found that the effects of the clad temperature and the power of the heater rods are small and the effects of the pressure and the inlet flow rate are large on the relationship between the superficial steam velocity and the void fraction in a bundle geometry during the reflood phase. It was, also, found that there is no distinct difference of the void fractions caused by the different flow patterns in the wet clad region and in the dry clad region in a bundle geometry during the reflood phase, when compared at the same superficial steam velocities.

Furthermore, the applicability of Cunningham-Yeh's void fraction correlation was investigated under a wide range of conditions anticipated during the reflood phase. The range of conditions under which Cunningham-Yeh's correlation predicts the void fraction within an error band of ±20% were made clear.     

Evaluation of a numeric procedure for flow simulation of a 5 × 5 PWR rod bundle with a mixing vane spacer

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.     

CFX simulation of high temperature thermal–chemical experiment: CS28-2

A computational fluid dynamics (CFD) model of a post-blowdown fuel channel analysis for aged CANDU reactors with crept pressure tube has been developed, and validated against a high temperature thermal–chemical experiment: CS28-2. The CS28-2 experiment is one of three series of experiments to simulate the thermal–chemical behavior of a 28-element fuel channel at a high temperature and a low steam flow rate which may occur in severe accident conditions such as a LBLOCA (large break loss of coolant accident) of CANDU reactors. Pursuant to the objective of this study, the current study has focused on understanding the involved phenomena such as the thermal radiation and convection heat transfer, and the high temperature zirconium-steam reaction in a multi-ring geometry. Therefore, a zirconium-steam oxidation model based on a parabolic rate law was implemented into the CFX-10 code, which is a commercial CFD code offered from ANSYS Inc., and other heat transfer mechanisms in the 28-element fuel channel were modeled by the original CFX-10 heat transfer packages. To assess the capability of the CFX-10 code to model the thermal–chemical behavior of the 28-element fuel channel, the measured temperatures of the fuel element simulators (FES) of three fuel rings in the test bundle and the pressure tube, and the hydrogen production in the CS28-2 experiment were compared with the CFX-10 predictions.     

A finite element LES for high-Re flow in a short-elbow pipe with undisturbed inlet velocity

The paper is concerned with a large-eddy simulation (LES) for a high-Reynolds-number flow in a short-elbow pipe, which can potentially be employed in the primary piping system of the Japan Sodium-cooled Fast Reactor (JSFR). The basic performance of the LES is studied for an elbow pipe flow without turbulence at inlet boundary at Re = 1.2 × 10⁶ by comparison with a flow observed in a 1/3-scale water experiment, where the flow disturbance at the pipe inlet is small. In setting up the computational conditions, special care was taken to ensure that the mesh subdivision was suitable for the simulation of the pipe flow through a theoretical consideration. We discuss the effects of the turbulence model (Smagorinsky model, WALE model) and the inlet velocity profile on the results. The mechanism of the pressure fluctuation and the origin of the fluid force are also discussed with the aid of spectral analysis and the visualization of essential hydraulic quantities.     

Analysis of low-heat-flux sodium boiling test in a 37-pin bundle by the two-fluid model computer code SABENA

This paper summarizes the development of a new detailed multi-dimensional multi-field computer code SABENA and its application to an out-of-pile low-heat-flux sodium boiling test in a 37-pin bundle. The semi-implicit numerical method employed in the two-fluid six-equation two-phase flow model has proved in solving a wide spectrum of sodium boiling transients in a rod bundle under low pressure conditions. The code is capable of predicting the spatial incoherency of the boiling, dryout on fuel cladding surfaces and fuel pin heat transfer. Essential to the successful application of such a mechanistic model computer code are validational efforts aimed at the LMFBR accident phenomenology analyses. Through the simulation of the natural circulation boiling conditions, this study provides a consistent analytical interpretation of the experimental data. The important influences of such parameters as the inlet flow restriction and bundle geometry have been examined through interpretations of two-phase flow analysis including considerations of the flow instability induced dryout mechanism.     

带格架四棒束超临界水流动传热数值分析

棒束内超临界水流动传热是超临界水堆堆芯热工水力研究的重要内容,但对其认识还十分有限。本文针对四棒束内超临界水的流动传热现象开展数值模拟,特别分析了定位格架对棒束通道内流动和传热的影响。结果表明,采用SSG湍流模型计算所得到的棒束壁面温度和实验结果吻合良好,定位格架的存在影响下游流体的速度分布,显著提高格架下游的传热特性,交混系数有大幅上升,使得加热棒周向壁面温度分布更加平均,最高温度出现位置发生改变。     

Modeling and validation of a mechanistic tool (MEFISTO) for the prediction of critical power in BWR fuel assemblies

Westinghouse is currently developing the MEFISTO code with the main goal to achieve fast, robust, practical and reliable prediction of steady-state dryout Critical Power in Boiling Water Reactor (BWR) fuel bundle based on a mechanistic approach. A computationally efficient simulation scheme was used to achieve this goal, where the code resolves all relevant field (drop, steam and multi-film) mass balance equations, within the annular flow region, at the sub-channel level while relying on a fast and robust two-phase (liquid/steam) sub-channel solution to provide the cross-flow information. The MEFISTO code can hence provide highly detailed solution of the multi-film flow in BWR fuel bundle while enhancing flexibility and reducing the computer time by an order of magnitude as compared to a standard three-field sub-channel analysis approach.Models for the numerical computation of the one-dimensional field flowrate distributions in an open channel (e.g. a sub-channel), including the numerical treatment of field cross-flows, part-length rods, spacers grids and post-dryout conditions are presented in this paper. The MEFISTO code is then applied to dryout prediction in BWR fuel bundle using VIPRE-W as a fast and robust two-phase sub-channel driver code. The dryout power is numerically predicted by iterating on the bundle power so that the minimum film flowrate in the bundle reaches the dryout criteria. Predicted dryout powers (including trends with flow, pressure, inlet subcooling and power distribution) and predicted dryout locations (both axial and radial) are compared to experimental results, using the entire Westinghouse SVEA-96 Optima3 dryout database, and are shown to yield excellent results.     

The effect of a CANDU fuel bundle geometry variation on thermalhydraulic performance

The effect of fuel bundle geometry has been evaluated using the subchannel code ASSERT to enhance the thermalhydraulic performance. Based on the configuration of a standard 37-element fuel bundle, the element diameter in each ring has been changed while those of other rings are kept as the original size. The dryout power of each element in a fuel bundle has been obtained for the modified fuel bundle and compared with that of a standard fuel bundle. The results obtained show that the modification of element diameter largely affects the thermal characteristics of the subchannel on the upper region of a modified element due to the buoyancy drift effect. The explicit effect of the element size on the dryout power of a fuel bundle is investigated by acquiring comprehensive numerical solutions. It is found that the dryout power of a fuel bundle has a tendency to increase as the element size decreases, except for the case where the outer ring is modified. The dependency of the transverse interchange model on the present results has been assessed by examining the dryout power of a bundle for different mixing coefficients and buoyancy drift models.     

On turbulence models for rod bundle flow computations

In commercial computational fluid dynamics codes there is more than one turbulence model built in. It is the user responsibility to choose one of those models, suitable for the problem studied. In the last decade, several computations were presented using computational fluid dynamics for the simulation of various problems of the nuclear industry. A common feature in a number of those simulations is that they were performed using the standard k–ε turbulence model without justifying the choice of the model. The simulation results were rarely satisfactory. In this paper, we shall consider the flow in a fuel rod bundle as a case study and discuss why the application of the standard k–ε model fails to give reasonable results in this situation. We also show that a turbulence model based on the Reynolds stress transport equations can provide qualitatively correct results. Generally, our aim is pedagogical, we would like to call the readers attention to the fact that turbulence models have to be selected based on theoretical considerations and/or adequate information obtained from measurements.