全长尺寸5×5格架棒束通道两相流动特性研究

采用两相计算流体动力学(CFD)分析的方法,对全长尺寸格架棒束通道内过冷沸腾两相流动进行了数值模拟。将模拟得到的棒束通道中心4个子通道的平均空泡份额与实验值进行对比发现,在高空泡份额区域与实验值符合较好;在低空泡份额区域,计算值略高于实验值。两相CFD方法模拟得到了棒束通道内空泡份额的详细分布,观察到格架上游空泡份额集中在加热棒的周围,但在格架下游,子通道中心的空泡份额增加,加热棒周围的空泡份额减小,间接地证明了格架对临界热流密度(CHF)的提升作用。     

RELAP5大空间自然对流管束传热模型改进及验证

基于经验证的单相和两相大空间自然对流管束传热模型,对RELAP5进行了改进,使得程序具备了模拟单相和两相大空间自然对流管束传热的能力。采用改进后的系统程序RELAP5和改进前的系统程序RELAP5对试验模拟体进行了对比计算,并采用试验数据对改进后的程序进行了验证,结果表明,改进后的系统程序计算结果与试验数据吻合较好。      

高通量工程试验堆RELAP5建模与试验评价

出口母管破口失水事故(LOCA)是高通量工程试验堆(HFETR)安全评价的重要始发事件之一,本文基于RELAP5程序,建立了HFETR的数值计算模型,模拟了HFETR的LOCA试验工况;通过手动全开HFETR除气系统DN50阀模拟出口母管失水试验,获得了反应堆进出口压力、容补器压力和破口流量的变化,并通过试验数据验证了RELAP5程序的计算结果合理性,结果表明:RELAP5计算结果和实验结果吻合较好,最大相对误差为7.4%,说明利用RELAP5程序模拟低温中压压水型研究堆的系统瞬变可行。     

用AC-600非能动余热排出系统实验评估RELAP5程序

利用RELAP5程序对先进堆二次侧非能动堆芯余热排出系统实验的瞬态过程进行数值模拟。在微循环启动,有注水的工况下,比较了RELAP5程序的计算结果和实验数据,计算结果与实验基本一致。由此可见,利用RELAP5程序分析此类问题是可行的。瞬态计算结果还为先进压水堆非能动余热排出系统的设计提供参考。     

RELAP5程序预测第一类密度波不稳定边界的能力验证

为估算低温核供热堆的第一类密度波不稳定（Type-I DWO）边界,以确定其微沸腾运行模式的参数区间,本文建立了低温核供热堆NHR200相似性实验回路HRTL200的RELAP5数值模型。通过对比模拟结果与实验结果,评价了RELAP5/MOD3.2程序模拟Type-I DWO的一般特性以及预测不稳定边界的能力,分析了进、出口阻力系数、相间摩擦对模拟结果的影响。结果表明,RELAP5程序模拟Type-I DWO 的一般特性与实验符合较好;运行压力不高于25 bar（1 bar=105 Pa)时,程序计算的不稳定边界的过冷度边界值与实验值偏差在3 K以内;运行压力大于30 bar时,采用准确的相间摩擦关系式可以改善预测结果。因此,选取与回路相匹配的相间摩擦关系式后,RELAP5程序可以用于模拟和预测Type-I DWO。      

AP1000核电厂二回路主管道双端断裂流体喷射力计算分析

AP1000是先进的第三代压水堆核电厂,为确保核电厂在事故工况下的安全性,需对二回路主管道发生双端断裂的工况进行研究。本文采用RELAP5/MOD3.4软件对核电厂二回路突发主管道双端断裂的事故工况进行了数值模拟,计算得到断裂后管道破口处的喷放流量、压强、空泡份额及喷射力等物理参数的变化特性,并将计算结果与ANSI 58.2简化计算方法的结果进行了比较分析。结果表明,RELAP5/MOD3.4计算所得的喷射力小于简化计算方法所得结果。本文分析结果为进行AP1000核电厂的破裂管道甩击防护提供了基础。     

RELAP5程序与三维时空中子动力学模型的耦合以及改进研究

引入堆芯物理计算的两群三维时空中子动力学模型,对RELAP5程序的点堆中子动力学模型进行了改进,同时设计了可视化界面,可方便地实现人.机交互操作.计算结果与实际应用表明,改进后的RELAP5程序计算功能和精度得到提高,使用更加方便,在核动力装置的仿真方面有很好的应用前景.     

PRS换热器蒸汽凝结换热模型研究

使用RELAP5/MOD3.2程序对某型核动力装置二次侧非能动余热排出系统(PRS)1:1实验装置进行稳态计算,一些工况下计算结果同实验结果偏差较大。研究了汽-液界面剪切应力及系统高压等条件对层流和湍流状态下竖直管内蒸汽凝结模型的影响,并对模型进行了改进。改进后的RELAP5程序对该系统1:1实验装置进行稳态和瞬态计算,计算结果同实验结果符合良好。     

热工水力系统程序中数值水锤问题研究

基于两流体六方程的热工水力系统程序在计算蒸汽即将从控制体中消失或水即将充满控制体工况时,由于空泡份额较小的两相混合物和纯液相之间可压缩性的不连续变化以及离散动量方程的离散方法,可能会出现虚拟的压力峰值,即数值水锤现象。本文以热工水力系统分析程序RELAP5为参考对数值水锤问题的缓解方案进行了分析研究,给出了详细的检测逻辑以及修正方案,并应用于普赖尔管问题和冷凝实验工况的计算分析。结果显示,数值水锤缓解方案的启用能够缓解两流体程序中针对该类问题由于数值方法带来的压力瞬态效应,从而能够明显地降低压力峰值,避免了严重扭曲瞬态解的出现。数值水锤缓解方案减缓这一虚拟压力峰值,有利于提高程序计算的稳定性;针对该问题此方法可为同类型系统程序的开发及模型优化提供参考。      

基于RELAP5和FLUENT的多尺度耦合分析方法研究

基于二次开发得到的铅冷快堆一维系统程序RELAP5_LEAD和三维计算流体力学程序FLUENT,利用动态链接库技术和FLUENT用户自定义函数,开发了多尺度耦合分析程序RELAP5/FLUENT。在单相范围内,分别利用耦合程序RELAP5/FLUENT开展简单铅冷串联管道的瞬态流动和传热模拟、简单铅冷闭式回路的瞬态流动模拟,并与RELAP5_LEAD计算结果开展Code-to-Code对比分析。研究结果表明,RELAP5/FLUENT计算结果与RELAP5_LEAD模拟结果吻合良好,耦合程序的开发取得了初步成功,可用于分析铅冷快堆堆内的复杂三维热工水力现象。     

A numerical technique for analysis of transient two-phase flow in a vertical tube using the Drift Flux Model

This paper presents a numerical solution of one-dimensional transient two-phase flow in a vertical channel using the Drift Flux Model (DFM). The DFM treats the two phases as a mixture, but allows slippage between the gas and the liquid phase. The DFM was used for the calculation of velocity and fraction of each phase, combined with the most relevant closure relationships models for condensation, wall evaporation, and phasic velocities. The solution of the three conservation equations for the mixture and a continuity equation for the gas phases is obtained by a semi-implicit numerical method. A finite volume method is used to discretize the governing equations on a staggered grid in the computational domain. Satisfactory agreement is shown between predicted void fraction, RELAP5 code and available experimental data under both transient and steady state conditions. Numerical solution was also obtained for a wide two-phase flow conditions: system pressure, surface heat flux, mass flow rate and inlet sub-cooling to check the model ability to predict void fraction accurately. It is concluded, therefore, that the DFM is able to predict void fraction in subcooled flow boiling with sufficient accuracy. For pressures lower than 30 bars, the DFM overestimated the void fraction in comparison with the experimental data by about 15%. The model requires less computational power to simulate than other approaches and has no limitations on the nodalization process for numerical stability. It is therefore expected that development of presented model will be useful for the assessment of experimental data, as well as performing pre-test numerical experimentation.     

Assessment of RELAP5/MOD3.2.2 Gamma against ABB Atom 3×3-Rod Bundle Reflooding Tests

The results of the ABB Atom 3×3-Rod Bundle Reflooding Tests were used for assessment of the reflooding model used in RELAP5/MOD3.2.2 Gamma version. The assessment calculations were performed using the default calculation model options implemented in the code.The tests were performed to investigate the effects of different spacer grid designs on heat transfer during the reflooding period of a pressurized water reactor loss-of-coolant accident (LOCA). The tests were conducted under low-pressure and low-flow (LPLF) conditions using a PWR-type 3×3-rod bundle with full-length indirectly electrically heated, stepped cosine axial power-shaped heater rods. Three different spacer grid configurations were studied: spacer grids without mixing vanes, mixing vane spacer grids, and mixing vane spacer grids together with intermediate flow mixers (IFM).A total of 36 tests with different spacer grid configurations were calculated. For two selected basic tests with non-mixing spacer grids an extended comparison of calculated and measured parameters is presented. The comparison of the predicted and measured maximal cladding temperatures and quench times, which are the most important parameters in licensing calculations, is presented for all the performed tests.The assessment calculations were preceded by nodalization, time step, and moving mesh studies.The RELAP5/MOD3.2.2 Gamma code was found to still have several deficiencies in the reflood model. The calculation results show a satisfactory agreement with experimental inner peak cladding temperature, however the predicted temperature turn-around times and quench times are significantly too short. The results also show a significant over-prediction of the reflood heat transfer and the vapour temperatures. The void profile downstream the quench front is not correctly predicted either. Finally, the present reflood model does not properly reflect the effects of spacer grids on the reflood heat transfer.In spite of these deficiencies the improvements incorporated into RELAP5/MOD3.2 by the Paul Scherrer Institute (PSI) eliminated the unphysical behaviors such as continuous cooling without clear turn-around temperature and no visible quenching phenomena, which were shown in the reflood calculations by means of the RELAP5/MOD3.1 code.     

Development of a new bubbly-slug interfacial friction model for RELAP5

The development of a new bubbly-slug interfacial friction model for the Pressurized Water Reactor (PWR) safety code RELAP5 is described. The model is based on a set of best-estimate void fraction correlations which cover the full range of geometries and flow conditions encountered in PWR safety analysis. By exploiting the relationship between void fraction and interfacial friction that exists for steady, fully developed flow conditions, the correlations are converted into effective interfacial friction coefficients that can be applied in the code for transient as well as steady-state conditions. Assessments against separate effects tests indicate that the new model is more accurate than the existing model in many situations, particularly rod bundle geometries, and should never be significantly less accurate. The model has been implemented in a local version of RELAP5/MOD2 and in a pre-release version of RELAP5/MOD3 at Idaho National Engineering Laboratory (INEL).     

State-of-the-art computer code RELAP5 validation with RBMK-related separate phenomena data

Validation of the RBMK model, developed by employing best estimate system computer code RELAP5 is performed by employing the data from NPPs operation or from integral and separate effects facilities.Validation of the models on the basis of separate phenomena is necessary to perform due to the fact that RELAP5 code has been developed for PWRs, which operate at different conditions (pressure, temperature, coolant void fraction, etc.) from RBMKs. In addition to that, there is a number of phenomena specific for RBMK type reactors (oscillatory flow rate behaviour in parallel channels, flow stagnation in channels, stratification in long horizontal piping, etc.), which have not been studied during RELAP5 validation for PWRs.In the paper, RELAP5 models for separate effects related to RBMK-1500 are presented and modelling of transients is performed. Obtained results are compared with experimental data.     

基于RELAP5与MELCOR联合分析方法的压水堆严重事故研究

针对严重事故的模拟研究,本文提出结合热工水力系统程序和严重事故一体化程序的分析方法,以典型三环路传统压水堆为对象,分别采用RELAP5和MELCOR程序建立模型,分析在全厂断电叠加汽动辅助给水泵失效事故下系统的瞬态响应。为了尽可能地利用RELAP5计算早期热工水力响应,同时保证严重事故计算结果的准确性,以MELCOR锆合金氧化模型开始工作温度的下限,即包壳温度达到1 100 K作为程序衔接准则并利用RELAP5的大编辑功能,提取所需计算结果导入MELCOR输入卡作为初始参数继续模拟。计算结果表明,数据连接过程整体保持了连续性,两种方法计算得出的主冷却剂系统压力、堆芯和稳压器水位、燃料包壳温度等参数的数值以及堆芯传热恶化和压力容器失效等现象的时序存在不同程度的差异,例如堆芯熔毁时间延后了约538 s。由于采用了RELAP5计算严重事故前的系统暂态响应,联合分析方法的计算结果比单独使用MELCOR分析的结果更加准确,该方法可以提高传统严重事故分析的可靠性。     

Effect of compensation error in drift-flux parameters on predictions of thermal-hydraulic parameters in nuclear safety system analysis codes

Void fraction in a nuclear reactor core is one of the most important parameters in a safety analysis using nuclear reactor thermal-hydraulics system analysis codes such as TRAC-BF1, RELAP5 and TRACE. Interfacial shear term governing void fraction in the two-fluid code is often estimated by Andersen approach which uses drift-flux type correlation to compute the interfacial shear term. The accuracy of two drift-flux parameters such as distribution parameter and drift velocity is anticipated to affect the accuracy of predicted void fraction significantly. In principle, the distribution parameter and drift velocity are independent parameters which should be determined by local gas and liquid velocities and void fraction. However, due to very limited local data, the distribution parameter and drift velocity are commonly determined by area-averaged void fraction and superficial gas and liquid velocities. This “approximate method” is acceptable when the distribution parameter and drift velocity are used together. However, in the Anderson's approach, the distribution parameter and drift velocity determined by the approximate method are used separately which may cause some compensation error in code calculations. In view of the great importance in accurately computing the interfacial shear term, the effect of the compensation error on the predicted void fraction is investigated. Intensive sensitivity analysis suggests the compensation error propagating to void fraction only up to 1% for steady state computations, whereas the effect of the compensation error on the predicted void fraction for transient computations becomes larger because temporal reduction of drag force may cause the increase in void fraction. A prototypic nuclear power plant analysis for ATWS scenario suggests that the overestimation of the void fraction may affect the neutron flux calculation.     

开式自然循环流动不稳定性的频域法分析

非能动安全壳热量导出系统依靠自然循环导出事故后排入安全壳内的热量,但在运行过程中也可能发生流动不稳定性现象。本文以某开式自然循环非能动安全壳热量导出系统为对象,建立了描述该系统行为的数学模型和本构关系,运用小扰动法对守恒方程进行线性化,通过Laplace变换获得系统质量流速随加热段进口焓变化的传递函数。基于Nyquist稳定性判据,分析了热工参数变化对该自然循环系统稳定性的影响。结果表明:系统的流动稳定性本质上受空泡份额随质量含气率的变化关系的制约,在一定范围内,随着质量含气率的增大,空泡份额对质量含气率的敏感性减弱,系统趋于稳定。     

基于RELAP5与MELCOR联合分析方法的压水堆严重事故研究

针对严重事故的模拟研究,本文提出结合热工水力系统程序和严重事故一体化程序的分析方法,以典型三环路传统压水堆为对象,分别采用RELAP5和MELCOR程序建立模型,分析在全厂断电叠加汽动辅助给水泵失效事故下系统的瞬态响应。为了尽可能地利用RELAP5计算早期热工水力响应,同时保证严重事故计算结果的准确性,以MELCOR锆合金氧化模型开始工作温度的下限,即包壳温度达到1 100 K作为程序衔接准则并利用RELAP5的大编辑功能,提取所需计算结果导入MELCOR输入卡作为初始参数继续模拟。计算结果表明,数据连接过程整体保持了连续性,两种方法计算得出的主冷却剂系统压力、堆芯和稳压器水位、燃料包壳温度等参数的数值以及堆芯传热恶化和压力容器失效等现象的时序存在不同程度的差异,例如堆芯熔毁时间延后了约538 s。由于采用了RELAP5计算严重事故前的系统暂态响应,联合分析方法的计算结果比单独使用MELCOR分析的结果更加准确,该方法可以提高传统严重事故分析的可靠性。     

基于RELAP5的氦氙流动换热计算模块开发与验证

反应堆热工系统分析程序是开展热工水力计算与安全评价的重要工具。为开发适用于氦氙气冷空间堆的热工系统分析程序,本文在RELAP5/MOD40程序中拓展了氦氙混合气体（He Xe）物性计算模块,添加了适用于He Xe的传热关系式,将拓展后程序计算值与实验值进行对比。结果表明：程序默认的Sutherlands定律用于He Xe物性计算时将引入较大误差;Dittus Bolter公式对He Xe对流换热时的Nu预测偏高,将导致不保守的壁温计算结果。拓展后的程序对He Xe压降和换热计算结果均与实验值吻合较好,验证了程序开发的正确性以及程序用于He Xe流动换热计算的功能。本研究可为系统层面程序开发奠定基础。