竖直环形狭缝通道内环状流的预测模型

对竖直环形狭缝通道内环状流流动沸腾传热理论模型进行了分析，以液膜质量、动量和能量守恒方程为基础，结合汽芯动量方程建立了竖直环形狭缝通道内环状流的数学物理模型。对该模型进行数值求解，得出了液膜厚度、液膜内的速度分布和温度分布、内—外管的换热系数以及通道内压降值，并与实验值进行了比较。     

雷诺数关系式对膜状凝结湍流区传热计算的影响

膜状凝结现象广泛存在于核电站安全壳和稳压器中。关于膜状凝结液膜湍流区的传热模型,目前未明确辨析基于质量和能量关系的两种雷诺数关系式的差别。本文针对管外纯蒸汽自然对流膜状冷凝传热,定量地分析雷诺数关系式对膜状凝结液膜湍流区传热计算的影响。基于液膜湍流区修正项的一般性假设,推导了膜状凝结湍流区传热系数的表达式。同时,分别与雷诺数关系式Remass和Reenergy联立,求解得到不同雷诺数关系式之间以及对应的膜状凝结传热系数之间的关系。分析表明:受普朗特数Pr的影响,在膜状凝结液膜湍流区,雷诺数关系式Remass和Reenergy差别明显,并存在关于Pr的分界点。基于Remass和Reenergy得到的膜状凝结平均传热系数及其相对偏差是Re和Pr的非线性函数。当0.1Pr4.0且Re1 600时,基于Reenergy和Remass得到的膜状凝结平均传热系数相对偏差在-60%和+60%之间。通过实验和理论验证,在膜状凝结液膜湍流区基于Reenergy得到的膜状凝结传热系数更加准确。     

非能动安全壳冷却系统膜状冷凝强化换热设计

非能动安全壳冷却系统(PCCS)能在反应堆发生事故时将安全壳内部的热量及时导出,避免安全壳因超温、超压而失效。为强化换热,本文设想在安全壳内部安装阻隔带和液滴收集装置,通过降低层流区液膜厚度、扰动不可凝气体隔离层并充分利用湍流的换热强化作用,降低总的换热热阻,提高换热效率。以AP1000为例,依托GDLM模型对改进前后安全壳的换热情况进行分析,结果表明,通过安装阻隔带和液滴收集装置,能降低安全壳壁面的液膜厚度,提高壁面热流量,从而实现强化换热。     

曲率对竖直向上环形狭窄通道内环状流特性的影响

基于三流体分离流模型，以液膜质量、动量和能量守恒方程为基础，结合汽芯动量方程，对双面加热垂直向上流动环形狭窄通道内环状流特性进行数值模拟。将计算结果与实验结果相比较，两者符合较好。通过数值模拟，分析了曲率对环状流特性的影响，得到了曲率对液膜厚度、液膜内温度、液膜内速度、临界热流密度等的影响曲线。曲率越大，内液膜越薄，而外液膜越厚。内管干涸时，临界热流密度随曲率的减小而增大；外管干涸时，则反之。     

环形狭缝通道内环状流模型的数值分析

对环形狭缝通道内的环状流建立了分离流模型。应用质量、动量和能量守恒方程 ,加上相应的边界条件和使方程组封闭的经验关系式 ,对环形狭缝通道的内、外液膜厚度、液膜内的速度分布和温度分布 ,以及内、外管的换热系数进行了数值计算求解     

不同壁面特性下液膜铺展性能数值模拟

液膜在安全壳表面的流动铺展性能对非能动安全壳冷却系统有重要的影响。采用三维数值模拟方法研究了降膜板表面形貌、接触角与液相雷诺数对液膜铺展性能的影响,并将结果与前人的实验和理论结果对比,吻合良好。研究发现,当降膜板为横向波纹板时液膜完全铺展时间、液膜厚度及界面湍动程度明显大于平板与纵向波纹板,此时波纹板波谷处会有循环流动产生。随接触角的增加或液相雷诺数的减小,液膜逐渐从完整流转变为片状流、溪流、滴状流。在纵向波纹板对液膜的导流与撕裂综合作用下,随接触角的变化,液膜的铺展性能与平板相比也发生较大的变化。     

PCS表面液膜覆盖率对安全壳完整性影响分析

钢制安全壳是防止严重事故工况下放射性物质向环境释放的最后一道屏障,因此有必要研究分析事故条件下安全壳外液膜覆盖率对安全壳完整性影响,以得到安全壳在事故工况下的失效裕度。应用非能动安全壳分析程序,建立了大功率非能动反应堆非能动安全壳冷却系统(Passive Containment Cooling System,PCS)的热工水力模型,并以冷段双端剪切事故为基准研究对象,分别研究了水分配器单一故障和出水管堵管叠加水分配器故障两种事故工况。分析结果表明,两种事故工况在液膜覆盖率大于35%时,均不会出现短期安全壳超压超温失效;事故后24 h,液膜覆盖率低于45%时,安全壳出现长期冷却失效。此次研究得出结论:在流量大于61.76 m3·h-1、安全壳液膜覆盖率大于45%时,事故发生后24 h安全壳不会失效。     

均匀加热窄通道内环状流液膜厚度和换热系数的预测模型

为探究窄矩形通道内环状流的流动传热特性,根据液膜的质量、动量和能量方程以及汽芯的动量方程建立了环状流的预测模型。对该模型进行数值求解,得出了窄矩形通道内环状流区域的沸腾换热系数,并分析了热流密度、质量流速和矩形通道尺寸对液膜厚度的影响。结果表明:该模型能很好地预测沸腾换热系数,其误差在±30%以内,且热流密度和矩形通道的尺寸对液膜厚度的影响效果比较大。     

UO_2芯块辐照密实、肿胀和裂变气体释放

本文从裂变气体扩散,气孔非平衡收缩、演化、迁移、聚合,气泡成核、长大等三个过程出发,导出气孔、气泡非平衡收缩、长大方程,气孔、气泡和气孔链分布函数的积分、微分方程。给出了辐照密实、肿胀、裂变气体释放率和瞬时释放表达式。     

用解析和数值方法研究动量淀积深度分布

基于解析和数值两种方法,作者研究了Glazov于1994年提出的计算由原子弹性碰撞引起动量淀积的空间分布.本文证明了Glazov推导出的积微分方程是不正确的,因为其中将两个原本独立的变量"纠缠"起来.分布函数在靶表面的"奇点"正是由这种"纠缠"引起的.动量淀积分布函数所满足的传统Boltzmann输运方程与Glazov方程并不等价.所以,由Glazov方程推导出的一切结果就都成问题了.另一方面,作者证明了由Glazov同样的高精度计算(通常n=300阶矩,28位小数)精度不足以再现Glazov文中的大部分图表.在这个工作中,作者用Padé逼近以及更高精度(n=980阶矩,500位小数)计算了动量淀积分布函数的富利叶变换f(k,η)和fL(k).我们的数值计算结果表明:对于这类输运问题的研究,Padé逼近是一种强大的理论工具.     

Experimental simulation of corium dispersion phenomena in direct containment heating

In a direct containment heating (DCH) accident scenario, the degree of corium dispersion is one of the most significant factors responsible for the reactor containment heating and pressurization. To study the mechanisms of the corium dispersion phenomenon, a DCH separate effect test facility of 1:10 linear scale for Zion PWR geometry is constructed. Experiments are carried out with air-water and air-woods metal simulating steam and molten core materials. The physical process of corium dispersion is studied in detail through various instruments, as well as with flow visualization at several locations. The accident transient begins with the liquid jet discharge at the bottom of the reactor pressure vessel. Once the jet impinges on the cavity bottom floor, it immediately spreads out and moves rapidly to the cavity exit as a film flow. Part of the discharged liquid flows out of the cavity before gas blowdown, and the rest is subjected to the entrainment process due to the high speed gas stream. The liquid film and droplet flows from the reactor cavity will then experience subcompartment trapping and re-entrainment. Consequently, the dispersed liquid droplets that follow the gas stream are transported into the containment atmosphere, resulting in containment heating and pressurization in the prototypic condition. Comprehensive measurements are obtained in this study, including the liquid jet velocity, liquid film thickness and velocity transients in the test cavity, gas velocity and velocity profile in the cavity, droplet size distribution and entrainment rate, and the fraction of dispersed liquid in the containment building. These data are of great importance for better understanding of the corium dispersion mechanisms.     

严重事故条件下安全壳响应模拟研究

压水堆核电厂发生严重事故期间，从主系统释放的蒸汽、氢气以及下封头失效后进入安全壳的堆芯熔融物均对安全壳的完整性构成威胁。以国内典型二代加压水堆为研究对象，采用MAAP程序进行安全壳响应分析。选取了两种典型的严重事故序列：热管段中破口叠加设备冷却水失效和再循环高压安注失效，堆芯因冷却不足升温熔化导致压力容器失效，熔融物与混凝土发生反应（MCCI），安全壳超压失效；冷管段大破口叠加再循环失效，安全壳内蒸汽不断聚集，发生超压失效。通过对两种事故工况的分析，证实了再循环高压安注、安全壳喷淋这两种缓解措施对保证安全壳完整性的重要作用。     

AP1000 ADS-4夹带卸压试验模拟分析

第4级自动降压系统（ADS-4）是AP1000极为重要的非能动安全设施。ADS-4能在AP1000小破口失水事故中为反应堆系统提供可控卸压。然而,大量的冷却剂可通过卸压过程中ADS-4夹带和上腔室夹带被带到安全壳中,从而引发堆芯裸露和堆芯熔化事故。为研究小破口事故中的ADS-4夹带卸压和上腔室夹带过程,在以AP1000为原型、按直径/高度比1∶5.6设计建造的ADS-4喷放卸压试验回路（ADETEL）中,研究了不同初始压力、压力容器混合液位和加热功率下的夹带和卸压行为,以及反应堆内部构件的夹带沉积效应。试验数据表明,大量的水在短时间内迅速通过ADS-4支管被夹带出来。液体的夹带率和压力容器混合液位的降低速率随系统初始压力的增加而增大。值得注意的是,在本试验特定工况下,初始压力为0.5 MPa时出现堆芯裸露。堆内构件对夹带量和压力容器混合液位无显著影响。     

Modeling pressure suppression pool hydrodynamics in the ABWR containment

This paper focuses on the assessment of pressure suppression pool hydrodynamics in the advanced boiling water reactor (ABWR) containment under design-basis, loss-of-coolant accident (LOCA) conditions. The paper presents a mechanistic model for predicting various suppression pool hydrodynamics parameters. A phenomena identification and ranking table (PIRT) applicable to the ABWR containment pool hydrodynamics analysis is used as a basis for the development of the model. The highly ranked phenomena are represented by analytic equations or empirical correlations. The best estimate and several sensitivity calculations are performed for the ABWR containment using this model. Results of the sensitivity calculations are also presented that demonstrate the influence of key model parameters and assumptions on the pool hydrodynamics parameters. A comparison of model predictions to the results of the licensing analyses shows reasonable agreement. Comparison of the results of the proposed model to experimental data shows that the model predicted top vent clearance time, the pool swell height, and the bubble breakthrough elevation are within 10% of the data. The predicted pool surface velocity and the liquid slug thickness are within 30% of the measurements, which is considered adequate given the large uncertainties in the experimental measurements.     

Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal–hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.     

Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

For the design of an LWR containment one of the important conditions to be considered is the rapid rise of internal pressure and temperature caused by a loss-of-coolant accident (LOCA) of the primary cooling system. The phenomena occurring within a containment during a LOCA are currently investigated through experiments with a model containment. The experimental results are compared with the results of model calculations to improve the calculational methods.An experimental facility was built, consisting of a primary coolant circuit and a special model containment. The model containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m³ and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross sections. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiments a PWR configuration with nine compartments has been installed. The model scales of the compartment volumes and the overflow areas are about 1:64 compared to the 1200 MW PWR plant Biblis A.Up to now the test facility has been used for four trial runs and nine PWR LOCA experiments with single- and double-ended pipe ruptures of 100 mm dia. in a steam generator compartment and in the nozzle compartment. The initial conditions of the pressurized water in the coolant circuit before rupture were 140 bar and 290°C. About 0.1 sec after the rupture the flow rate at the site of rupture reaches its maximum of about 400 kg/sec (single-ended rupture) and 800 kg/sec (double-ended rupture). From the compartment where the rupture takes place a water-steam-air mixture streams through openings into the other compartments of the containment. Differential pressures between the compartments were measured with maximums of up to a few bar 0.15–0.5 sec after rupture, depending on the positions of rooms and transducers.Approximately 30–40 sec after rupture the blowdown has finished and the pressure in the containment has reached about 4–5 bar. The maximum pressure in a model containment is lower and the decrease of the pressure by condensation is faster than in a full-scale containment, due to the greater ratio of inner surface area to volume of a model containment. During blowdown the temperature of the containment atmosphere rises to about 150°C. Several minutes later the temperature of the concrete walls has increased non-uniformly causing considerable stress in the walls. Approximately 30 min after rupture measurements on the outside of the outer containment wall show a temperature-caused strain of about 30–60% of the maximum pressure-caused strain. A comparison between experiments and calculations shows discrepancies indicating the need for further development of calculational methods.     

基于威斯康辛冷凝试验本体的结构改进及分析

先进压水堆(APWR)是第三代核电技术的代表堆型之一,它采用了非能动安全系统,提高了安全性能。非能动安全壳冷却系统(PCCS)主要利用蒸汽的冷凝来带走安全壳内的热量。本文主要介绍了威斯康辛大学进行的冷凝试验的试验本体结构,应用ANSYS软件对其结构进行了应力分析,并在现有结构的基础上对外部加强筋布置进行了一定的改进和优化。通过计算和比较可以看出,经过改进后的加强筋布置,不仅满足原有的试验要求,结构布置合理,更提高了试验本体的承压能力,使其能够满足更高试验压力的需要。