基于流动与传热网络的HTR-PM堆内热工水力模拟

球床模块式高温气冷堆核电站（HTR-PM）全尺寸模拟机的开发是示范工程的重要组成部分,HTR-PM堆内热工水力过程的模拟是模拟机开发的关键技术之一。本文针对堆内热工水力过程的模拟进行了研究,根据堆内几何结构和热工水力过程的特点,采用组件搭建的方式建立了HTR-PM堆内流动与传热过程的计算模型,基于所建立的流动与传热网络模拟方法,编制了相应的模拟计算程序,实现了HTR-PM堆内热工水力过程的模拟,给出了反应堆50%FP、100%FP稳态工况、控制棒误提升事故工况的模拟结果,通过与设计分析程序THERMIX的比较进行了验证。对比结果表明,模拟方法和基于流动与传热网络的计算模型能够满足模拟机的开发要求,反映了堆内热工水力过程的特点。     

倾斜与摇摆条件下一体化反应堆自然循环特性研究

通过建立海洋条件下的附加力模型与控制体空间坐标求解模型,开发了基于RELAP5/MOD3.1程序的海洋条件热工水力分析程序。研究了海洋条件下一体化反应堆IP200的自然循环特性,分析倾斜与摇摆条件对自然循环的影响。计算结果表明:倾斜会降低堆芯流量,导致左右侧环路冷却剂流量不一致,影响直流蒸汽发生器的换热特性;摇摆情形下,环路的附加压降主要由切向力贡献;摇摆轴偏离中心位置以及倾斜和摇摆的叠加运动均会打破环路间的热工水力对称性,增大堆芯流量的波动幅度。     

超临界水冷堆燃料棒性能分析程序适用性研究

根据超临界水冷堆(SCWR)燃料棒的热工水力特点,基于压水堆(PWR)燃料棒性能分析程序的理论模型和计算方法研究燃料包壳的物性模型和超临界水(SCW)与燃料包壳的传热模型,建立适用于SCWR燃料棒的性能分析程序——SCWRFPA。采用SCWRFPA和可分析SCWR的热工水力子通道程序ATHAS分别对1/8欧洲超临界轻水堆(HPLWR)燃料组件燃料棒进行计算,其计算结果基本一致。     

堵流条件下矩形并联通道非均匀流动传热模拟方法研究

为建立矩形并联通道非均匀流动传热模拟方法,针对板型燃料元件的安全分析提供新的模拟方法和工具,本研究采用一维两流体模型和燃料元件二维导热模型开发热工水力瞬态分析程序,对堵流条件下非均匀流动传热进行模拟。通过数值模拟得到不同堵流工况下流量分配和燃料温度分布,此外对4种不同功率分布下燃料元件二维导热效应进行研究。研究结果表明,堵流后并联通道流量和传热量将重新分配,二维导热模型使燃料元件截面温度场分布更均匀。本文开发的热工水力瞬态分析程序能够用于板型燃料元件非均匀流动传热现象的模拟。     

海洋条件下反应堆热工水力参数的子通道计算

针对海洋条件下反应堆的子通道热工水力分析,建立了海洋运动附加力模型和瞬态入口边界,将起伏、摇摆及复合运动的附加力关系式用于子通道模型的轴向和横向动量方程,并应用到COBRAⅢC程序将其改造为适应海洋条件的反应堆子通道分析程序。作为验证,计算了加热实验通道和"奥陆"堆在起伏运动情况下热通道的临界热流密度比(CHFR)、出口空泡份额和冷却剂流量,并与文献结果对比。还详细计算了"奥陆"堆在起伏、不同摇摆中心和复合运动情况下,热通道的CHFR和不同位置子通道出口的热工水力参数。研究表明:海洋条件下反应堆的子通道热工水力参数随运动呈周期性变化;起伏运动对子通道的压降影响较大,摇摆运动对子通道冷却剂的流量和温度影响较大。     

钠冷快堆乏燃料组件热工水力分析程序开发

为了对示范快堆乏燃料组件的热工水力特性进行分析,自主研发了钠冷快堆乏燃料组件热工水力分析程序SPATANS。该程序基于子通道分析方法,采用适用于低流量下的流动换热和交混关系式。针对乏燃料组件棒束区进行计算,得到组件不同高度处各子通道的温度、压力等热工参数,并将计算结果与三维计算流体力学FLUENT程序的结果进行对比分析。结果表明:自主研发程序的计算结果与FLUENT程序的计算结果较为吻合,偏差在工程可接受范围内,且其计算效率明显高于FLUENT程序。初步表明SPATANS程序可用于钠冷快堆乏燃料组件热工水力分析,并具有良好的应用前景。     

钠冷快堆乏燃料组件热工水力分析程序开发

为了对示范快堆乏燃料组件的热工水力特性进行分析，自主研发了钠冷快堆乏燃料组件热工水力分析程序SPATANS。该程序基于子通道分析方法，采用适用于低流量下的流动换热和交混关系式。针对乏燃料组件棒束区进行计算，得到组件不同高度处各子通道的温度、压力等热工参数，并将计算结果与三维计算流体力学FLUENT程序的结果进行对比分析。结果表明：自主研发程序的计算结果与FLUENT程序的计算结果较为吻合，偏差在工程可接受范围内，且其计算效率明显高于FLUENT程序。初步表明SPATANS程序可用于钠冷快堆乏燃料组件热工水力分析，并具有良好的应用前景。     

秦山核电厂主蒸汽管道破裂事故的RELAP5／MOD2分析

通过使用RELAP5/MOD2程序对秦山核电厂主蒸汽管道破裂事故的计算,对该程序的临界流模型和传热模型进行分析,并与其它大型热工水力分析程序的计算结果及实验结果进行比较。在计算过程中,对RELAP5/MOD2程序汽水分离器模型的使用进行修正,使之符合核电厂安全评审计算的要求。     

应用于反应堆热工水力程序的核态沸腾传热关系式评价

本文以反应堆热工水力分析程序COSINE开发为背景,针对燃料棒和冷却剂换热及压力容器外部冷却时的核态沸腾两种特殊的工况,研究常用于计算热工水力程序的核态沸腾传热关系式的计算结果随影响参数的变化关系,比较不同范围内各关系式计算结果的差异程度和敏感性,为程序中用户选项的设置和进一步实验验证提供参考意见,研究表明高过热度工况最需进行实验验证,反应堆热工水力分析程序计算这两种工况下的核态沸腾传热更适宜选用Chen、Schrock-Grossman1、Wright和SchrockGrossman2公式。     

高性能轻水堆中子物理-热工水力耦合行为分析

针对1种典型的"三流程"超临界水堆——高性能轻水堆(HPLWR)开发了中子物理-热工水力耦合分析程序,并对其堆芯进行了核热耦合计算。基于该程序开展了传热关系式敏感性研究,得出适用于HPLWR核热耦合的传热关系式,进而对HPLWR进行中子物理-热工水力耦合行为计算,得出了一些关键参数沿轴向的分布规律。结果表明:开发的程序可较好地分析高性能轻水堆的中子物理-热工水力耦合行为。     

The development and validation of a thermal hydraulic code in rolling motion

An advanced thermal hydraulic code is established on the basis of RELAP5/MOD3.3 code for the investigation of the thermal hydraulic behavior of nuclear power systems. The RELAP5 code is modified by adding a module calculating the effect of rolling motion and introducing new flow and heat transfer models. The experimental data are used to validate the theoretical models and calculation results. It is shown that the advanced flow and heat transfer models could correctly predict the frictional resistance and heat transfer coefficients in rolling motion. The thermal hydraulic code is used to simulate the operation of a natural circulation system in rolling motion. The calculation results are in good agreement with experimental data. The relative discrepancies between calculation results and experimental data are less than 5%.     

The influence of ocean conditions on thermal-hydraulic characteristics of a passive residual heat removal system

There are many differences between the flow and heat transfer characteristics of nuclear reactors under ocean and land-based conditions for the effects of ocean waves. In this paper, thermal hydraulic characteristics of a passive residual heat removal system (PRHRS) for an integrated pressurized water reactor (IPWR) in ocean environment were investigated theoretically. A series of reasonable theoretical models for a PRHRS in an IPWR were established. These models mainly include the core, once-through steam generator, nitrogen pressurizer, main coolant pump, flow and heat transfer and ocean motion models. The flow and heat transfer models are suitable for the core with plate-type fuel element and the once-through steam generator with annular channel, respectively. A transient analysis code in FORTRAN 90 format has been developed to analyze the thermal–hydraulic characteristics of the PRHRS under ocean conditions. The code was implemented to analyze the effects of different ocean motions on the transient thermal-hydraulic characteristics of PRHRS. It is found that the oscillating amplitudes and periods of the system parameters are determined by those of the ocean motions. The effect of rolling motion is more obvious than that of pitching motion when the amplitudes and periods of rolling and pitching motions are the same. The obtained analysis results are significant to the improvement design of the PRHRS and the safety operation of the IPWR.     

Thermal-hydraulic behavior of a marine reactor during oscillations

The effect of ship motion, such as heaving and rolling, on the thermal-hydraulic behavior of marine reactors was investigated. The COBRA-IV-I CODE was modified to analyse the thermal-hydraulic performance on the critical heat flux under oscillating acceleration conditions. The critical heat flux in the code was verified experimentally using freon as a comparison. The Critical Heat Flux Ratio (CHFR) at the hottest channel of the PWR subchannel was analysed using the same code. A system code RETRAN-02/MOD2-GRAV was developed by improving RETRAN-02/MOD2 to simulate the thermal hydraulic transient under ship motion. It was verified by comparison using the experimental results of both two-phase natural circulation flow under heaving motion and single-phase natural circulation flow at an inclined attitude. The code was used to analyse reactor plant behavior in the nuclear ship Mutsu. Natural circulation flow during rolling motion was investigated experimentally. The characteristics of loop flow and core flow rates were clarified. The core flow rate correlated well with the Reynolds number of rolling motion.     

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (I)

A computer code SEFDAN is developed for one-dimensional thermal-hydraulics in a partially uncovered core of a light water reactor during a severe core damage accident. The developed models include:

1. Froth level (or dry-out level) calculation

2. Transition and mixing between convection flow regimes in convective heat transfer

3. Radiant heat transfer between solid walls and flowing gas

4. Heat generation by zirconium-water reaction

5. Crucibilization effect of zirconium-oxide layer

6. Steam starvation effect on zirconium-water reaction.

This code does not calculate motion of fuel rod material but predicts the beginning of relocation. The major affecting models, froth level calculation model, heat transfer model and crucibilization model, are verified through analyses of experiments. This code can be used for thermal hydraulic analysis of a severe accident and fuel damage experiment until significant material relocation occurs.     

两流体模型在环形窄缝内的应用研究

采用EPRI最新开发的Chexal-Harrison相壁相间摩擦模型和简化的相壁相间传热模型,构造了适用于环形窄缝内沸腾传热和流动的两流体模型,并编制了热工水力计算程序——THYME程序.与实验数据比较,分析了环形窄缝套管在不同负荷下Relap5/Mod3.2程序和本文程序的计算结果.计算结果表明,Relap5/Mod3.2低估了环形蒸发管的蒸汽温度,本文计算结果与实验数据较为一致.     

Unsteady Reynolds Averaged Navier–Stokes simulation of the turbulent flow pulsation and coherent structures in a tight lattice in rolling motion

The flow in a tight lattice is strongly affected by the quasi-periodic lateral flow pulsations caused by large scale vortices. This kind of large scale vortices is largely responsible for the momentum and heat exchange across the gaps. In rolling motion, the coherent structure and flow oscillation are affected by an additional force. The coherent structure in rolling motion is more significant than that in no rolling motion. The oscillation period in rolling motion is about 10% bigger than that in no rolling motion. The rolling motion can affect the coherent structure. However, the effect of rolling motion on the thermal hydraulic parameters, i.e. wall temperature and bulk temperature, is very limited. The wall temperature and wall shear stress in rolling motion and no rolling motion are nearly the same. The additional force due to rolling motion can change the moving characteristics of coherent structures, but its effect on the turbulent flow and heat transfer is weak.     

摇摆条件下子通道内湍流流体的数值分析

利用FLUENT软件分析了摇摆条件对典型四棒束间的湍流流体流动和传热特性的影响机理。摇摆运动会对棒束间流体的流动传热特性产生一定影响。RSM模型可以很好地描述摇摆条件下子通道内的参数分布。摇摆周期变化带来的径向附加力的变化不会对摩擦阻力系数、传热系数和Reynolds应力产生影响。在摇摆条件下,摩擦阻力系数、传热系数和Reynolds应力呈周期性变化,但最大摩擦阻力系数所在时刻并不固定,而最大传热系数却始终是在流速最大的时刻。     

摇摆条件下典型通道间湍流的流动传热特性

利用Fluent软件分析了摇摆条件对典型四棒束间的湍流流体流动和传热特性的影响机理。摇摆运动会对棒束间流体的流动传热特性产生一定影响,但不会对绝热通道与加热通道内流体流动相似性产生影响。而当摇摆幅度较大时,径向附加力会使通道横截面上的参数分布发生显著的变化,进而影响流体的流动与传热特性。在摇摆条件下,随着P/D（棒间距/棒直径）的逐渐减小,尤其是小于1.1时,典型棒束间流体的流动传热特性发生明显变化。     

CFD analysis of flow and heat transfer of turbulent pulsating flow in a tube in rolling motion

The investigation of flow and heat transfer of turbulent pulsating flow is of vital importance to the nuclear reactor thermal hydraulic analysis in ocean environment. In this paper, the flow and heat transfer of turbulent pulsating flow is analyzed. The calculation results are firstly verified with experimental data. The agreement between them is satisfactory. The effect of spanwise and wall-normal additional forces is significant in small Reynolds number, and decreases with Reynolds number increasing. The rolling axis and rolling radius contribute slight to the flow and heat transfer. The effect of velocity oscillation period on the heat transfer is limited than that of Reynolds number and oscillating velocity Reynolds number. The traditional empirical correlations could not predict the flow and heat transfer of turbulent pulsating flow in rolling motion.     

华龙一号非能动安全壳热量导出系统热工水力特性研究

本文针对华龙一号非能动安全壳热量导出系统（PCS）,基于漂移流模型开发了一套一维自然循环瞬态计算程序。利用该程序对PCS内热工水力特性进行了分析研究,得到PCS自然循环流量、换热系数、换热器进出口温度、上升管路竖直段出口含气率及水箱水位等热工水力参数随PCS换热功率的变化。本文研究结果将为评估华龙一号PCS的换热能力提供可靠工具,对PCS的设计和改进也具有指导意义,并为后续开发能够模拟带有PCS的安全壳内热工水力行为的程序打下基础。