ADS-4空气-硼酸夹带试验研究

为研究真实工况下的ADS-4夹带现象,以CAP1400为原型按1∶1的比例设计搭建了FATE试验台架。硼酸溶液用来模拟反应堆堆芯中真实工况的流体。利用数据采集系统和高速摄像仪记录夹带过程,将所得的试验数据与纯水工况和已有的模型进行比较。结果表明：对于夹带起始和稳态夹带率,硼酸工况和纯水工况的试验结果与之前的模型均不同。硼酸工况和纯水工况的现象基本相似,弗劳德数随夹带起始液位的增加而减小。两者的夹带起始均有回滞效应,且自上而下的夹带起始更易发生。硼酸工况和纯水工况之间也存在着差异,这主要是由于两者的物理特性参数不同而造成的,尤其是密度和黏度的影响。     

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。     

先进非能动核电厂DEDVI事故热工水力模拟分析

采用Relap5/Mod3.4程序建立了先进非能动核电厂的事故分析模型,包括反应堆冷却剂系统(RCS)、简化的二回路系统和专设安全设施。针对小破口失水事故(SBL,OCA)中的直接安注管双端断裂事故(DEDVI)进行分析,并着重对SBLOCA现象识别和排序表(PIRT)中对其影响较大的液滴夹带进行敏感性分析。分析结果表明,对直接安注管双端断裂事故,破口和自动卸压系统(ADS)能够有效地使反应堆冷却剂系统降压,堆芯补水箱(CMT)、安注箱(ACC)和安全壳内置换料水箱(IRWST)能够迅速实现堆芯补水,确保堆芯冷却。对液滴夹带的敏感性分析表明,对于位置较高的第4级ADS,喷放流量对液滴夹带模型比较敏感,使用均相流模型计算时,其液相流量显著高于非均相流模型。     

SCDAP/Relap5 3.2程序分层夹带模型敏感性分析

在AP1000中,连接堆芯补水箱和冷腿间的压力平衡管线中的气泡份额决定了堆芯补水箱的注入量,其中,气泡源自冷腿中的分层夹带。为研究AP1000核电站中气-液分层夹带现象对堆芯非能动余热排出系统的整体特性的影响,本文以Relap5/Mod3.2作为计算平台,建立了AP1000小破口失水事故模型并进行了数值计算,对比了采用与不采用水平分层夹带模型的计算结果,发现该模型对事故发展有重要的影响。     

DVI管小破口失水事故实验研究

在模块化小型反应堆非能动安全系统综合模拟实验装置上进行了压力容器直接注入（DVI）管小破口失水事故实验,研究了DVI管小破口失水事故过程中的热工水力现象和非能动安全系统运行特性。研究结果表明：模块化小型反应堆DVI管小破口失水事故中,非能动安全系统可对堆芯进行注水,有效导出堆芯衰变热量,保护堆芯安全。     

波动管小破口失水事故实验研究

在模块化小型反应堆非能动安全系统综合模拟实验装置上进行波动管小破口尺寸失水事故实验,研究波动管小破口失水事故过程中的热工水力现象和非能动安全系统运行特性。模块化小型反应堆发生失水事故后,压力平衡管和安注管线内流体的密度差可以驱动堆芯补水箱(CMT)内的冷流体注入反应堆压力容器,压力平衡管裸露后CMT安注流量出现波动;安注箱(ACC)的安注对事故初期的堆芯冷却效果显著;经自动卸压系统卸压后,内置换料水箱(IRWST)可以对堆芯进行持续稳定的安注和冷却。研究结果表明:波动管小破口失水事故中,非能动安注系统可以对堆芯进行有效注水,并带走堆芯衰变热量。     

AP1000中ADS-4第4级自动降压系统夹带试验研究

为研究AP1000反应堆第4级自动降压系统(ADS-4)夹带卸压过程,以AP1000核电厂为原型,按1:5.6模化比例设计建造了ADS-4喷放卸压试验回路(ADETEL)。用高速摄像仪记录试验现象并进行了详细分析。将试验数据和现有试验数据和模型进行对比,并得出以下结论:当热管段内液位较低时,夹带量会随着热管段内液位降低而迅速减小;在热管段内流动情况相同(流动参数相同)情况下,夹带起始在,竖直支管和水平主管管径比较小的工况下更容易发生;在相同的热管段相对液位下,AP1000中ADS-4支管内液体的夹带率较AP600低。     

AC600全压堆芯补水箱补水实验研究

全压堆芯冰箱（ＣＭＴ）是ＡＣ６００压水堆非能动高压安注系统的主要设备。全压堆芯补水箱补水实验主要研究中,小破口失水事故时ＣＭＴ的重力排放特性,为验证安全分析计算机程序试验数据,中国核动力研究院建造了ＣＭＴ补水实验装置,并在该装置上模拟反应堆主管道中,小破口失水事故动态工况,完成了ＣＭＴ补水实验,本文给出了小破口失水事故工况堆芯水箱补水试验结果与分析。,     

美国第六次反应堆失水试验

【英国《应用原子新闻》1980年11月11日报道】据美国核管理委员会最近报道,美国能源部在爱达荷国立工程实验室进行了第六次反应堆失水试验。这次试验模拟了反应堆堆芯给水系统中一个小管子破裂后主冷却剂泵停转时发生的事件。     

美国第六次反应堆失水试验

【英国《应用原子新闻》1980年11月11日报道】据美国核管理委员会最近报道,美国能源部在爱达荷国立工程实验室进行了第六次反应堆失水试验。这次试验模拟了反应堆堆芯给水系统中一个小管子破裂后主冷却剂泵停转时发生的事件。     

竖直向下大、小支管气相夹带起始可视化研究

已有气相夹带起始模型均是基于竖直向下的小支管夹带所得到的,用于分析大支管气相夹带时并不适用。本文基于可视化实验,对竖直向下的大、小尺寸T型管的气相夹带起始点进行对比研究。选择与主管道直径比为0.625和0.1的大、小支管,并采用有机玻璃作为实验管道材料,空气和水为实验工质。其实验现象表明：大、小支管夹带起始均有漩涡,且漩涡现象大致相同,但在大支管条件下会出现气室,夹带气体进入支管后也会在气室下翻腾,且在相同液相折算速度下,大支管夹带起始液位会高于小支管。     

大型先进压水堆ADS-4夹带试验研究

针对大型先进压水堆的ADS-4夹带现象,按照实际比例1∶1搭建了全尺寸ADS-4试验台架(FATE),在不同工况下进行了多组夹带试验。将试验数据与不同模型的结果进行对比,并将试验数据进行拟合,得到夹带起始的模型关系式。试验结果显示：拟真实工况与纯水工况的现象较相似,支管弗劳德数随夹带起始气腔高度的增加而增加;在不同工况的组合中均出现了回滞现象;由于黏性和表面张力的作用,拟真实工况的夹带起始相比纯水工况更难发生。无论是纯水工况、硼酸工况还是杂质工况,除少部分试验数据外,大部分稳态夹带率的试验数据与Welter模型存在较大差异,相对误差均大于20%。     

New Theoretical Model for Two-Phase Flow Discharged from Stratified Two-Phase Region through Small Break

A theoretical and experimental study was conducted to understand two-phase flow discharged from a stratified two-phase region through a small break. This problem is important for an analysis of a small break loss-of-coolant accident (LOCA) in a light water reactor (LWR). The present theoretical results show that a break quality is a function of h/h_b , where h is the elevation difference between a bulk water level in the upstream region and the break and b the suffix for entrainment initiation. This result is consistent with existing experimental results in literature. An air-water experiment was also conducted changing a break orientation as an experimental parameter to develop and assess the model. Comparisons between the model and the experimental results show that the present model can satisfactorily predict the flow rate and the quality at the break without using any adjusting constant when liquid entrainment occurs in a stratified two-phase region. When gas entrainment occurs, the experimental data are correlated well by using a single empirical constant.     

Assessment of an entrainment model in annular-mist flow for a three-field TRAC-M

A three-field model for annular-mist flow regime was developed and directly implemented into the one-dimensional components of TRAC-M/F90. The three-field model implemented in the present study was firstly developed based on the physical models of COBRA-TF.In the development of the present three-field model, the empirical correlations including entrainment and de-entrainment models in the literature were investigated and tested with the three-field model implemented in TRAC-M/F90. The entrainment model was assessed using Kataoka and Ishii's and Würtz's models.The use of Würtz's entrainment model shows excellent agreement with high pressure, steam/water experiments, while considerable scattering was observed in low pressure, steam/water and low pressure, air/water experiments. Kataoka and Ishii's model was found to be the best application in air/water, porous-sinter cases. Because of the narrowness of the applicability of Kataoka and Ishii's model, Würtz's model was selected as the base model for entrainment rate.The final combination of the empirical correlations is suggested and validated against steam–water experiments of various conditions. The agreements with experimental data were at 25.0% RMS error.     

液膜蒸干模型在液态金属CHF预测中的应用

为对圆管中环状流烧干型沸腾临界进行预测,建立了考虑液滴沉积夹带作用的液膜蒸干模型。沉积率、夹带率等相关关系式采用水等常规流体的已有关系式。在一定参数范围内,将模型预测结果与实验数据和经验关系式进行了比较。结果表明：基于常规流体的液膜蒸干模型大体可应用于液态金属,但在临界热流密度（CHF）较大时可能造成较大偏差;钠、钾两种液态金属在CHF较低时预测准确度区别不大,但在CHF较高时区别较为明显。为更加准确预测液态金属CHF,应开发专门的沉积率、夹带率等相关关系式。     

微细化沸腾传热实验研究

气泡微细化沸腾是沸腾到达某个临界热负荷后,加热面温度升高不大,与该临界热负荷相比,热流密度大幅提高的沸腾现象。本文在设计完成一可视化实验装置的基础上,通过高速摄影仪观察并结合采集的壁温数据,对常压下直径为10 mm铜加热面上的池式气泡微细化沸腾现象进行了研究,并讨论了液体过冷度对其的影响。实验发现,气泡微细化沸腾状态下,加热面上生成1层极其不稳定的气膜,气液交界面上不停地有大量微小气泡生成并以极高速度射入过冷液体中。随加热面热流密度的增大,气膜厚度波动周期缩短,气膜最大厚度减小,所生成微小气泡的直径也明显减小。实验中获得的最高热流密度达9 MW/m²。     

Film splitting model with entrainment

Critical heat flux (CHF) is one of the major parameters for calculating safety margin of nuclear power plant systems based on US-Nuclear Regulatory Commission's safety regulations (defined under 10CFR50-46). Exceeding the CHF point suddenly increases the fuel clad temperature so that it may result in damage to the nuclear fuel pins. Dryout is a special CHF physical phenomenon occurred in Boiling Water Reactors (BWRs). Firstly, the present work identifies several existing dryout correlations and notes the major assumptions for the derivation of them. It follows with the derivation of a film splitting model (originally proposed by Chun et al., in 2003) for annular geometry where the liquid film is broken into an inner and outer film. Film splitting can result in better prediction of the critical film thickness where dryout can occur due to local variations especially due to turbulence and other effects. Secondly, an alternate approach to solving the film splitting model of Chun et al. is employed to evaluate the model against the same experimental data. It was observed that liquid entrainment and deposition into and out of the film layer were the missing components in the model and the model predictions shown in Chun et al. (2003) so that the model predictions with alternative solution methods were compared against experimental data by excluding entrainment/de-entrainment phenomena. Thirdly, the film split model is updated by adding entrainment/de-entrainment models to make the film splitting model more realistic. Fourthly, new film splitting model is compared against both tubular and pipe data. It was observed that entrainment and deposition are significant contributors in the new mechanistic dryout model. The mass flux of the inner film is mostly overestimated while the mass flux of the outer film is overestimated for some of the experimental cases when compared to the original correlation.     

Sodium Mist Behavior in Cover Gas Space of LMFBR Out-Pile Experiment

Sodium mist behavior in argon cover gas space over sodium pool was investigated experimentally using of a test vessel having cover gas volume of ～100l. Mass concentration and gravitational settling flux of the mist (i.e. sodium aerosol) were determined between pool temperature range of 290～520°C. Apparatuses used for the concentration determination were a He-Ne laser mist concentration meter (LCM) and mist traps, which were developed specifically for the present experiment. The mist gravitational settling flux φ was determined by using of collection plates, which were exposed to the mist entrained cover gas over sodium pool.

The experimental results revealed that the concentration C was very high from 0.15 to 20g/m³. It increased with pool temperature. The mist particle radii, which were determined from the gravitational settling flux φ and decay curve of the concentration, were from 1.5 to 13 μm in the range of the present experiment. The vapor evaporation rate φe from the pool surface, which were also determined from the experimental φ data, were found to increase with saturated vapor pressure P_s at pool surface. This rate did not depend so strongly on experimental geometries and cover gas flow convection pattern over sodium pool. Evaluations of transient C changing and steady state C fluctuation indicated that the mist formation was occurred near the pool surface.