Comparison of thermo-hydraulic performances of large scale vortex flow (LSVF) and small scale vortex flow (SSVF) mixing vanes in 17 × 17 nuclear rod bundle

Spacer grids in the nuclear fuel rod assembly maintain a constant distance between rods, secure flow passage and prevent the damage of the rod bundle from flow-induced vibration. The mixing vanes attached to the spacer grids generate vortex flows in the subchannels and enhance the heat transfer performance of the rod bundle. Various types of mixing vanes have been developed to produce cross flows between subchannels as well as vortex flows in the subchannels.The shapes of the mixing vane have been improved to generate larger turbulence and cross flow mixing. In the present study, two types of large scale vortex flow (LSVF) mixing vanes and two types of small scale vortex flow (SSVF) mixing vanes are examined. SSVF-single is conventional split type and SSVF-couple is split type with different arraying method. LSVF mixing vane has different geometry and arraying method to make large scale vortex. 17 × 17 rod bundle with eight spans of mixing vanes is simulated using the IBM 690 supercomputer. The FLUENT code and IBM supercomputer is employed to calculate the flow field and heat transfer in the subchannels.Turbulence intensities, maximum surface temperatures of the rod bundle, heat transfer coefficients and pressure drops of the four kinds of mixing vanes are compared. LSVF mixing vanes produced higher turbulence intensity and heat transfer coefficient than SSVF mixing vanes. Consequently, LSVF mixing vane increases the thermal efficiency and safety of the rod bundle.     

CHF experiment and CFD analysis in a 2 × 3 rod bundle with mixing vane

In this study, the CHF enhancement using various mixing vanes is evaluated and the flow characteristics are investigated through the CHF experiments and CFD analysis.CHF tests were performed using 2 × 2 and 2 × 3 rod bundles and with R-134a as the working fluid. The test section geometry was identical to that of commercial PWR fuel assembly not including the heated length (1.125 m) and number of fuel rods. From the CHF tests, it was found that the CHF enhancement using mixing vanes under higher mass flux (1400 kg/m² s) and lower pressure (15 bar) conditions is larger than the CHF enhancements under other conditions. Among the mixing vanes used in this study, the swirl vane showed the best performance under relatively low pressure (15 bar) and mass flux (300-1000 kg/m² s) conditions and the hybrid vane performed best near the PWR operating conditions.The detailed flow characteristics were also investigated by CFD analysis using the same conditions as the CHF tests. To calculate the subcooled boiling flow, the wall partitioning model was applied to the wall boundary and various two-phase parameters were also considered. The reliability of the CFD analysis in the boiling analysis was confirmed by comparing the average void fractions of the analysis and the experiments: the results agreed well. From the CFD analysis, the void fraction flattening as a result of the lateral velocity induced by the mixing vane was observed. By the lateral motion of the liquid, the void fraction in the near wall was decreased and that of the core region was increased resulting in the void fraction flattening. The decrease of the void fraction in the near wall region promoted liquid supply to the wall and consequently the CHF increased. For the quantification of the void flatness, an index was developed and the applicability of the index in the CHF assessment was confirmed.     

Single-phase CFD applicability for estimating fluid hot-spot locations in a 5 × 5 fuel rod bundle

High-thermal performance PWR spacer grids require both of low pressure loss and high critical heat flux (CHF) properties. Therefore, a numerical study using computational fluid dynamics (CFD) was carried out to estimate pressure loss in strap and mixing vane structures. Moreover, a CFD simulation under single-phase flow condition was conducted for one specific condition in a water departure from nucleate boiling (DNB) test to examine the applicability of the CFD model for predicting the CHF rod position. Energy flux around the rod surface in a water DNB test is the sum of the intrinsic energy flux from a rod and the extrinsic energy flux from other rods, and increments of the enthalpy and decrements of flow velocity near the rod surface are assumed to affect CHF performance. CFD makes it possible to model the complicated flow field consisting of a spacer grid and a rod bundle and evaluate the local velocity and enthalpy distribution around the rod surface, which are assumed to determine the initial conditions for the two-phase structure. The results of this study indicate that single-phase CFD can play a significant role in designing PWR spacer grids for improved CHF performance.     

A second order turbulence model based on a Reynolds stress approach for two-phase boiling flow and application to fuel assembly analysis

High-thermal performance PWR (pressurized water reactor) spacer grids require both low-pressure loss and high critical heat flux (CHF) properties. Numerical investigations on the effect of angles and position of mixing vanes and to understand in more details the main physical phenomena (wall boiling, entrainment of bubbles in the wakes, recondensation) are required.In the field of fuel assembly analysis or design by means of CFD codes, the overwhelming majority of the studies are carried out using two-equation Eddy Viscosity Models (EVM), especially the standard K-? model, while the use of Reynolds Stress Transport Models (RSTM) remains exceptional.The simulation of swirling flow generated by the mixing vanes plays an important role for the prediction of the CHF for the fuel assemblies. For this reason, according to [14] and [Mimouni et al., 2009b], rotation effects and RSTM model are more specifically addressed in the paper.Before comparing performance of EVM and RSTM models on fuel assembly geometry, we performed calculations with simpler geometries, the DEBORA case and the ASU-annular channel case. ASU-annular channel case has already been addressed in [14] and [Mimouni et al., 2009b].Then, a geometry closer to actual fuel assemblies is considered. It consists of a rectangular test section in which a 2 × 2 rod bundle equipped with a simple spacer grid with mixing vanes is inserted. The influence of the turbulence model on target variables linked to CHF limitation will be discussed. Moreover, the sensitivity to the mesh refinement will be particularly examined. The study of this case is a further step towards the modelling of the two-phase boiling flow in real-life grids and rod bundles.     

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.     

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.     

Phenomenological investigations on the turbulent flow structures in a rod bundle array with mixing devices

Detailed turbulent flow profiles have been measured on a square sub-channel geometry with typical mixing devices. For a fine examination of the lateral flow structure on a sub-channel geometry with 2D LDA, a 5 × 5 rod bundle array was fabricated as 2.6 times larger than the real bundle size. The mixing devices used were a typical split type and a swirl type. The experiments were performed at the condition of Re = 48,000 (axial bulk velocity 1.48 m/s) and the water loop was maintained at the conditions of 35 °C and 1.5 bar during an operation. As for the results, distinct intrinsic flow features were observed according to the type of mixing devices. In a typical split type, there was no remarkable swirling flow within a sub-channel and the lateral flow was vigorous in the gaps. In the swirl type, a single swirling flow was dominant within a sub-channel and there were relatively small lateral flows in the gaps.     

Experimental study on the Reynolds number dependence of turbulent mixing in a rod bundle

An experimental study for Reynolds number dependence of the turbulent mixing between fuel-bundle subchannels, was performed. The measurements were done on a triangular array bundle with a 1.20 pitch to diameter relation and 10 mm rod diameter, in a low-pressure water loop, at Reynolds numbers between 1.4 × 10³ and 1.3 × 10⁵.The high accuracy of the results was obtained by improving a thermal tracing technique recently developed. The Reynolds exponent on the mixing rate correlation was obtained with two-digit accuracy for Reynolds numbers greater than 3 × 10³. It was also found a marked increase in the mixing rate for lower Reynolds numbers.The weak theoretical base of the accepted Reynolds dependence was pointed out in light of the later findings, as well as its ambiguous supporting experimental data.The present results also provide indirect information about dominant large scale flow pulsations at different flow regimes.     

An experimental study on the critical heat flux for low flow of water in a non-uniformly heated vertical rod bundle over a wide range of pressure conditions

An experimental study of the critical heat flux (CHF) has been performed for a water flow in a non-uniformly heated vertical 3 × 3 rod bundle under low flow and a wide range of pressure conditions. The experiment was especially focused on the parametric trends of the CHF and the applicability of the conventional CHF correlations to a return-to-power conditions of a main steam line break accident whose conditions might be a low mass flux, intermediate pressure, and a high inlet subcooling. The effects of the mass flux and pressure on the CHF are relatively large and complicated in the low pressure conditions. At a high mass flux or a low critical quality, the local heat flux at the CHF location sharply decreases with an increasing local critical quality. However, at a low mass flux or a high critical quality, the local heat flux at the CHF location shows a nearly constant value regardless of the increase of the critical quality. The CHF data at the very low mass flux conditions are correlated well by the churn-to-annular flow transition criterion or the flow reversal phenomena. Several conventional CHF correlations predict the present return-to-power CHF data with reasonable accuracies. However, the prediction capabilities become worse in a very low mass flux of below about 100 kg/(m² s).     

One-dimensional critical heat flux concerning surface orientation and gap size effects

Tests were conducted to examine the critical heat flux (CHF) on a one-dimensional downward heating rectangular channel having a narrow gap by changing the orientation of the copper test heater assembly in a pool of saturated water under atmospheric pressure. The test parameters include both the gap sizes of 1, 2, 5 and 10 mm, and the surface orientation angles from the downward-facing position (180°) to the vertical position (90°), respectively. Also, the CHF experiments were performed for pool boiling with varying heater surface orientations in the unconfined space at atmospheric pressure using the rectangular test section. It was observed that the CHF generally decreases as the surface inclination angle increases and as the gap size decreases. In consistency with several studies reported in the literature, it was found that there exists a transition angle at which the CHF changes with a rapid slope. An engineering correlation is developed for the CHF during natural convective boiling in the inclined, confined rectangular channels with the aid of dimensional analysis. This correlation agrees with the experimental data of this study within ±20%.     

带搅混格架5×5棒束流场的激光多普勒测量

采用激光多普勒测速（LDV）系统,对带有2道搅混格架的5×5棒束格架下游4个横截面的流场进行了轴向流速测量。实验测得了各个截面上的平均轴向速度分布和轴向脉动速度均方根（RMS）分布。通过比较不同截面的平均速度与RMS速度的差异,分析了定位格架下游流场的演变规律;比较了2道格架下游的实验数据,分析了上游流场对格架效应的影响。本文实验数据是在充分重复性实验的基础上获得的,可以为计算流体动力学（CFD）结果的验证和评价提供基准参考。      

Flooding limited CHF in a vertical 3 × 3 rod bundle with non-uniform axial heat flux

KAERI has performed an experimental study on the critical heat flux (CHF) under zero flow conditions with a non-uniformly heated 3 × 3 rod bundle. Experimental conditions are in the range of a system pressure from 0.50 to 14.96 MPa and inlet water subcooling enthalpies from 68 to 352 kJ/kg. The test section used in the present experiments consisted of a vertical flow channel, upper and lower plenums, and a non-uniformly heated 3 × 3 rod bundle. The experimental results show that the CHFs in low-pressure conditions are somewhat scattered within a narrow range. As the system pressure increases, however, the CHFs show a good parametric trend. The CHFs occur in the upper region of the heated section, but the locations of the detected CHFs move gradually in a downward direction with the increase of the system pressure. Even though the effects of the inlet water subcooling enthalpies and system pressure of the flooding CHF are relatively smaller than those of the flow boiling CHF, the CHF increases by increasing the inlet water subcooling enthalpies. Several existing correlations for the countercurrent flooding CHF based on Wallis's flooding correlation and Kutateladze's criterion for the onset of flooding are compared with the CHF data obtained in the present experiments to examine the applicability of the correlations.     

反应堆棒束通道搅混翼数值研究

定位格架上的搅混翼是核反应堆燃料组件中的关键部件,其性能对棒束通道热工水力特性有重要的影响。以带单层定位格架的5×5棒束为研究对象,对搅混翼排布方式及末端形状对格架下游的流场和温度场的影响进行数值模拟研究。计算结果表明,改变搅混翼的排布方式,压降几乎不受影响,但格架下游流场和传热情况却因排布方式的改变而发生显著变化;将搅混翼末端形状改为弧形,压降较典型撕裂型搅混翼没有明显差异,但换热情况得到明显改善。      

Experimental and numerical simulation of air-water two-phase flow in the rod bundle with grid spacer

1 Introduction Grid spacer is the key part of reactor fuel assem-bly. The presence of spacers in fuel assemblies affectsvarious thermal-hydraulic characteristics of the reactorcore. The grid spacer with fine performance can im-prove thermal-hydraulic performance of the core fuelassembly and enhance the critical heat flux withouttoo much augment of the pressure loss. As a result,the implementation of grid spacer with high thermalperformance provides more thermal margin, then in-creases s…     

Fluid-to-fluid modeling of critical heat flux in 4 × 4 rod bundles

Fluid-to-fluid modeling of critical heat flux (CHF) is to simulate the CHF behaviors for water by employing low cost modeling fluid, and the flow scaling factor is the key to apply the technique to fuel bundles. The CHF experiments in 4×4 rod bundles have been carried out in Freon-12 loop in equivalent nuclear reactor water conditions (P=10.0–16.0 MPa, G=488.0–2100.0 kg/m² s, X_cr=−0.20–0.30). The models in fluid-to-fluid modeling of CHF is verified by the CHF data for Freon-12 obtained in the experiment and the CHF correlation for water obtained by Nuclear Power Institute of China (NPIC) in the same 4×4 rod bundles. It has been found that the S.Y. Ahmad Compensation Distortion model, the Lu Zhongqi model, the Groeneveld model and Stevens–Kirby model overpredict the bundles CHF values for water. Then an empirical correlation of flow scaling factor is proposed. Comparison of the CHF data in two kinds of test sections for Freon-12, in which the distance of the last grid away the end of heated length is different, shows that the spacer grid, which is located at 20 mm away from the end of the heated length, has evidently influenced on the CHF value in the 4×4 rod bundles for Freon-12. This is different from that for water, and the need for further work is required.     

棒束通道定位格架沸腾临界特性数值研究

为研究核反应堆中定位格架及搅混翼对沸腾临界现象产生的影响,本文采用计算流体动力学（CFD）分析方法,探讨了棒束通道中定位格架的数目、位置和搅混翼的角度对于沸腾临界现象的影响。结果表明:定位格架会对主流流动产生阻力,同时定位格架数目越多,沸腾临界发生的温度也越高,但将定位格架布置在沸腾临界发生位置时,则可有效改善壁面传热环境并降低沸腾临界发生时的峰值温度。搅混翼的存在则会有效降低加热面附近空泡份额,改善传热环境,但搅混翼角度过大时会导致沸腾临界提前发生。      

Mass velocity and cold-wall effects on critical heat flux in an advanced light water reactor

The characteristics of Critical Heat Flux (CHF) were investigated for a square array of rod bundles which could possibly be loaded into an integral-type advanced light water reactor. The parametric effects of the mass velocity and the unheated rod were examined by conducting CHF experiments with 5 × 5 test bundles in a Freon-loop. The influence of a cold wall on the CHF was interpreted by introducing a simple phenomenological model which accounts for the influence of a thermal mixing inside the boiling channel. A local parameter CHF correlation applicable to an integral-type reactor was developed from the CHF data base for square-arrayed rod bundles. The local thermal–hydraulic conditions calculated by the subchannel analysis code MATRA were used for the optimization of the correlation coefficients. Correction factors for the low mass velocity, spacer grids, and the non-uniform axial power shapes have been devised which reflected the results of the data assessment and the experimental observations. As a result of the thermal margin evaluation at steady state conditions, it was revealed that the integral-type reactor core has a greater DNBR margin than a typical 1000 MW_e PWR core.     

Critical heat flux during natural convective boiling in inclined tubes submerged in saturated liquids

An experimental study was carried out to improve and expand understanding of boiling phenomena and the critical heat flux (CHF) during natural convective boiling in uniformly heated inclined tubes submerged in a pool of saturated liquids under atmospheric pressure. The test conditions were as follows: inter diameters of the test tubes ranged from 0.9 to 8.0 mm; heated lengths ranged from 100 to 400 mm, and inclination angles varied from 30° to vertical position. The test fluids were water and R-11. The experimental results showed that the CHF decreases with the increasing ratio of the tube length to the tube diameter, and with the reducing of the inclination angle. A semi-theoretical correlation, which originally used for the CHF during natural convective boiling in vertical tubes, was modified to predict the CHF occurs in the inclined tubes. The modified correlation agreed reasonably well with the present experimental data and other CHF data for narrow inclined annular tubes.     

燃料组件精细化定位格架模型开发及评价

为提高燃料组件子通道内两相局部参数预测的准确性,本文基于分布式阻力方法建立精细化定位格架模型,选用合适的摩擦阻力表达式,对格架上的交混翼进行精细化建模,采用Carlucci湍流交混模型计算湍流交混速率,引入阻塞因子计算由定位格架引起的湍流交混效应,并将建立的精细化定位格架模型植入子通道分析程序（ATHAS）,对压水堆子通道和棒束实验（PSBT）基准题进行计算分析。结果表明,本文开发的精细化定位格架模型能够提高燃料组件子通道内空泡份额和温度分布的预测准确性,为棒束通道流场、焓场计算和临界热流密度（CHF）预测奠定了基础。