Prediction of Transient Boiling Transition in BWR Fuel Rod Bundles by Subchannel Analysis Program MENUETT

Transient CHF (critical heat flux) tests of a 4 X 4 rod bundle were analyzed by the subchannel analysis program MENUETT. MENUETT is based on a non-equilibrium, five equation, two-phase flow model and is available both for steady state and transient analyses. Turbulent mixing and void drift effects are taken into account to calculate cross flows in fuel rod bundles. The tendency of calculated subchannel mass fluxes and qualities agreed with experimental data. By using a critical quality correlation obtained from steady state CHF data, the position of the earliest boiling transition could be predicted regardless of non-uniform axial heat flux distributions. This transition occurrence time was predicted within a difference of 0.1～0.3 s from the experimental time. MENUETT applicability was confirmed for transient calculations predicting thermalhydraulic behavior in bundles.     

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.     

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.     

Method of Critical Power Prediction Based on Film Flow Model Coupled with Subchannel Analysis

A new method was developed to predict critical powers for a wide variety of BWR fuel bundle designs. This method couples subchannel analysis with a liquid film flow model, instead of taking the conventional way which couples subchannel analysis with critical heat flux correlations. Flow and quality distributions in a bundle are estimated by the subchannel analysis. Using these distributions, film flow rates along fuel rods are then calculated with the film flow model. Dryout is assumed to occur where one of the film flows disappears. This method is expected to give much better adaptability to variations in geometry, heat flux, flow rate and quality distributions than the conventional methods.

In order to verify the method, critical power data under BWR conditions were analyzed. Measured and calculated critical powers agreed to within ±7%. Furthermore critical power data for a tight-latticed bundle obtained by LeTourneau et al. were compared with critical powers calculated by the present method and two conventional methods, CISE correlation and subchannel analysis coupled with the CISE correlation. It was confirmed that the present method can predict critical powers more accurately than the conventional methods.     

Analytical prediction of turbulent friction factor for a rod bundle

An analytical calculation has been performed to predict the turbulent friction factor in a rod bundle. For each subchannel constituting a rod bundle, the geometry parameters are analytically derived by integrating the law of the wall over each subchannel with the consideration of a local shear stress distribution. The correlation equations for a local shear stress distribution are supplied from a numerical simulation for each subchannel. The explicit effect of a subchannel shape on the geometry parameter and the friction factor is reported. The friction factor of a corner subchannel converges to a constant value, while the friction factor of a central subchannel steadily increases with a rod distance ratio. The analysis for a rod bundle shows that the friction factor of a rod bundle is largely affected by the characteristics of each subchannel constituting a rod bundle. The present analytic calculations well predict the experimental results from the literature with rod bundles in circular, hexagonal, and square channels.     

A critical heat flux approach for square rod bundles using the 1995 Groeneveld CHF table and bundle data of heat transfer research facility

The critical heat flux (CHF) approach using CHF look-up tables has become a widely accepted CHF prediction technique. In these approaches, the CHF tables are developed based mostly on the data bank for flow in circular tubes. A set of correction factors was proposed by Groeneveld et al. [Groeneveld, D.C., Cheng, S.C., Doan, T., 1986. 1986 AECL-UO Critical Heat Flux lookup table. Heat Transf. Eng. 7(1–2), 46] to extend the application of the CHF table to other flow situations including flow in rod bundles. The proposed correction factors are based on a limited amount of data not specified in the original paper. The CHF approach of Groeneveld and co-workers is extensively used in the thermal hydraulic analysis of nuclear reactors. In 1996, Groeneveld et al. proposed a new CHF table to predict CHF in circular tubes [Groeneveld, D.C., et al., 1996. The 1995 look-up table for Critical Heat Flux. Nucl. Eng. Des. 163(1), 23]. In the present study, a set of correction factors is developed to extend the applicability of the new CHF table to flow in rod bundles of square array. The correction factors are developed by minimizing the statistical parameters of the ratio of the measured and predicted bundle CHF data from the Heat Transfer Research Facility. The proposed correction factors include: the hydraulic diameter factor (K_hy), the bundle factor (K_bf), the heated length factor (K_hl), the grid spacer factor (K_sp), the axial flux distribution factors (K_nu), the cold wall factor (K_cw) and the radial power distribution factor (K_rp). The value of constants in these correction factors is different when the heat balance method (HBM) and direct substitution method (DSM) are adopted to predict the experimental results of HTRF. With the 1995 Groeneveld CHF Table and the proposed correction factors, the average relative error is 0.1 and 0.0% for HBM and DSM, respectively, and the root mean square (RMS) error is 31.7% in DSM and 17.7% in HBM for 9852 square array data points of HTRF.     

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。      

Pressure drop and velocity distribution in rod bundles with spacer grids

The paper contains experimental data and analysis of the pressure drop of turbulent flow through rod bundles. For laminar flow the dependence of the pressure drop on the pitch-to-diameter and wall-to-diameter ratios is discussed on the basis of theoretical analysis. In addition, correlations for the calculation of the pressure loss due to spacer grids are presented and compared with experimental data.Detailed measurements of the velocity distribution in a full bundle of 19 rods are compared with predictions for fully developed turbulent flow. Moreover, detailed measurements of the velocity distributions upstream and downstream of spacer grids typical for LMFBRs are discussed together with the mass flow separation and redistribution between the subchannels. The mass flow distribution found experimentally is compared with the predictions by a subchannel code. The status of experimental knowledge is shown.     

Critical heat flux prediction in rod bundles under upward low mass flux densities

The analysis of experimental data and results of calculations for heat transfer crisis in heated channels under low upward coolant mass flux densities is presented. This analysis allows the determination of the basic features of the boiling crisis phenomenon. It is shown that the methods currently used for critical heat flux (CHF) prediction have insufficient accuracy in the given range of parameters. A new relationship for the CHF calculation is presented. It should be used for the water–water energy reactor (WWER) and uran–graphite channel reactor—Chernobyl-type (RBMK) rod bundles, and is verified by the test data. The comparison of results obtained by a new CHF correlation and the relationship used in RELAP5/MOD3.1 Code is presented. It is shown that the latter overpredicts the CHF values at atmospheric pressure and for x_cr>0.4 and does not provide conservative estimations for the RBMK fuel bundles.     

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.     

Assessment of the interchannel mixing model with a subchannel analysis code for BWR and PWR conditions

The influence of the interchannel mixing model employed in a traditional subchannel analysis code was investigated in this study, specifically on the analysis of the enthalpy distribution and critical heat flux (CHF) in rod bundles in BWR and PWR conditions. The equal-volume-exchange turbulent mixing and void drift model (EVVD) was embodied to the COBRA-IV-I code. An optimized model of the void drift coefficient has been devised in this study as the result of the assessment with the two-phase flow distribution data for the general electric (GE) 9-rod and Ispra 16-rod test bundles. The influence of the subchannel analysis model on the analysis of CHF was examined by evaluating the CHF test data in rod bundles representing PWR and BWR conditions. The CHFR margins of typical light water nuclear reactor (LWR) cores were evaluated by considering the influence on the local parameter CHF correlation and the hot channel analysis result. It appeared that the interchannel mixing model has an important effect upon the analysis of CHFR margin for BWR conditions.     

Calculated heat transfer development in bundles

In many cases heat transfer in rod bundles can be considered as a superposition of several simple heat transports. A number of practical problems can thus be solved if solutions for these elementary transport are available. Two elementary heat transports in rod bundle geometry are investigated, namely the transport from a fuel rod surface into the adjacent subchannel and the transport from one subchannel into the next one. The first transport is characterized in terms of the Nusselt or Stanton numbers, while the latter in terms of the Stanton gap number or mixing factor. Hydraulically developed flow is assumed, with no feedback of heat transport on the flow condition. A three-dimensional numerical calculation by means of the finite difference method is applied to determine the eigenvalue solution of the response on a step change in heating for both cases. The method is tested on an internally heated concentrical annulus. The result is compared with available experimental and theoretical predictions. It is found that the heat transfer between subchannels is developed considerably slower in comparison with the development of the heat transport from fuel rod to subchannel.     

测量和预测棒束CHF数据对比方法研究

通过采用子通道程序FLICA建模分析5×5棒束临界热流密度试验,并分别采用直接代入法(DSM)和能量平衡法(HBM)两种方法利用已有临界热流密度关系式获得计算的临界热流密度,并将计算的临界热流密度与试验获得的临界热流密度对比分析,探讨了棒束临界热流密度试验数据的处理方法。结果表明,在棒束临界热流密度试验数据与已有关系式计算数据的对比中,HBM是一种更合理的方法。     

FIDAS: Detailed subchannel analysis code based on the three-fluid and three-field model

The Film Dryout Analysis Code in Subchannels, FIDAS, has been developed with the main objective of predicting dryout and post-dryout heat transfer in a channel and in rod bundles. In FIDAS, two-phase flow consisting of continuous liquid film, continuous vapor and entrained droplets is modeled by a three-fluid, three-field representation of 12 field equations, i.e. three continuity, three energy and six momentum equations. FIDAS can predict dryout without any empirical CHF correlations by introducing annular flow modeling and the ‘film dryout criterion’. Experiments on film flow characteristics, subchannel flow and enthalpy distributions, dryout and post-dryout heat transfer in tubes and rod bundles were analyzed to demonstrate the performance of FIDAS. The predictions of FIDAS are in close agreement with the experiments.     

Prediction of the single-phase turbulent mixing rate between two parallel subchannels using a subchannel geometry factor

This paper presents a simple method for predicting the single-phase turbulent mixing rate between adjacent subchannels in nuclear fuel bundles. In this method, the mixing rate is computed as the sum of the two components of turbulent diffusion and convective transfer. Of these, the turbulent diffusion component is calculated using a newly defined subchannel geometry factor F^* and the mean turbulent diffusivity for each subchannel which is computed from Elder's equation. The convective transfer component is evaluated from a mixing Stanton number correlation obtained empirically in this study. In order to confirm the validity of the proposed method, experimental data on turbulent mixing rate were obtained using a tracer technique under adiabatic conditions with three test channels, each consisting of two subchannels. The range of Reynolds number covered was 5000–66 000. From comparisons of the predicted turbulent mixing rates with the experimental data of other investigators as well as the authors, it has been confirmed that the proposed method can predict the data in a range of gap clearance to rod diameter ratio of 0.02–0.4 within about ±25% for square array bundles and about ±35% for triangular array bundles.     

CHF Experiments under Steady-State and Transient Conditions for Tight Lattice Core with Non-Uniform Axial Power Distribution

Abstract

Steady-state and transient critical heat flux (CHF) experiments were performed using triangular pitched 7-rod assemblies with non-uniform axial power distributions under the maximum pressure of 15.5 MPa. The onset of steady-state CHF was predicted within the uncertainty of 10% with the KfK correlation using the local flow conditions calculated by the subchannel analysis code COBRA-IV-T. On the other hand, various mechanistic CHF models did not agree with the steady-state CHF data. The transient CHFs under the conditions of flow reduction, power increase or flow and power simultaneous variation were predicted with the quasi-steady-state method within approximately the same uncertainty as the steady-state CHF experiments. The predictive capability did not depend on the transient speed within 30%/s of the flow reduction rate and within 120%/s of the power increase rate. It was also revealed that there exists large CHF margins under the thermal hydraulic conditions simulating the locked rotor accident and the control rod cluster ejection accident of the double-flat-core type high conversion pressurized water reactor (HCPWR).     

Check of the 2005-look up table for prediction of CHF in bundles

The check whether it is possible to use the 2005-look up table primary designed for heated pipes also for heated rod bundles gives the surprising result that the bundle critical power for five data sets of three different bundles and different power distributions are predicted by a simple method described above using the 2005-look up table within the accuracy reported by the authors of this table.     

The structure of turbulent flow through a wall subchannel of a rod bundle

An experimental investigation was performed to establish reliable information on the transport properties of turbulent flow through subchannels of rod bundles. Detailed data were measured of the distributions of the time-mean velocity, the turbulence intensities in all directions and hence, the kinetic energy of turbulence, of the shear stresses in the directions normal and parallel to the walls and of the wall shear stresses for a wall subchannel of a rod bundle of four parallel rods. The pitch to diameter ratio of the rods equal to the wall to diameter ratio was 1.07, the Reynolds number of this investigation was Re = 8.7 × 10⁴.On the basis of the data measured the eddy viscosities in the directions normal and parallel to the walls were calculated. Thus, detailed data of the eddy viscosities in direction parallel to the walls in rod bundles were obtained for the first time. The experimental results were compared with predictions by the VELASCO code. There are considerable differences between calculated and measured data of the time-mean velocity and the wall shear stresses. Attempts to adjust the VELASCO code against the measurements were not successful. The reasons of the discrepancies are discussed.     

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).     

环形通道内棒偏心和弯曲工况临界热流密度机理研究

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