The effects of break location on PWR small break LOCA: Experimental study at the ROSA-IV LSTF

This paper discusses the effect of break location on the break flow rate and break flow quality transitions during a small-break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). Results from five experiments conducted at the ROSA-IV Large Scale Test Facility (LSTF) are compared for this purpose. These experiments simulated a 2-inch break at the lower plenum, upper head, pressurizer top, cold leg, and hot leg, respectively. The controlling phenomena for the break flow quality transitions in cold-leg and hot-leg break experiments are described.     

基于CFD的蒸汽发生器U型管内回流现象数值分析

反应堆一回路系统在自然循环条件下，蒸汽发生器（SG）部分U型管内可能会出现回流现象，利用计算流体动力学（CFD）方法，对某非能动三代反应堆蒸汽发生器U型管内流体的流动传热特性进行数值模拟分析。选取6组不同管长的U型管，对比分析U型管内单相流体的流动传热特性。基于数值仿真结果，得出6组U型管质量流量-进出口压降曲线，并?T分析了U型管长度和一次侧进口流体温度与二次侧壁面温度温差（?T）对流体回流的影响。研究结果表明，当?T一定时，随着进出口压降的降低，长管内更容易发生回流。当U型管长度一定时，?T越小越容易发生回流。      

The results of 5% small-break LOCA tests and natural circulation tests at the ROSA-IV LSTF

Five 5% small-break loss-of-coolant accident (SBLOCA) experiments and two natural circulation experiments were conducted at the ROSA-IV Large Scale Test Facility (LSTF). The liquid holdup in the upflow side of steam generator (SG) U-tubes temporarily depressed the core collapsed liquid level below the bottom of core during the loop seal clearing in the cold-leg break SBLOCA tests. This phenomena was affected by the core power and core bypass but was affected little by the actuation of the high pressure injection system. Overall thermal-hydraulic phenomena in a loop seal line break test was similar to that of cold-leg break tests, however, the liquid holdup phenomena played a little role. In a hot-leg break test a temporary but rapid depression of the core liquid level was observed immediately after the initiation of accumulator injection which caused condensation and depressurization in the cold leg. The change of natural circulation flow rate with the decrease of primary system mass inventory was qualitatively the same as observed in Semiscale, LOBI and PKL. The SG effective overall heat transfer coefficient below the secondary-side collapsed liquid level was weakly dependent on the secondary side liquid level and the core power. The measured minimum heat transfer coefficient was 1.7 kW/m²K for the full secondary side mass inventory.     

Thermal–hydraulic characteristics of a next-generation reactor relying on steam generator secondary side cooling for primary depressurization and long-term passive core cooling

System experiments were conducted at the ROSA-V Large Scale Test Facility (LSTF) for investigation of new safety systems to mitigate consequences of postulated accidents in pressurized water rectors (PWRs). Tested systems included a steam generator (SG) secondary-side automatic depressurization system (SADS) and gravity-driven injection system (GDIS), which are candidates of safety systems for some next-generation PWR designs. The experimental results showed several thermal–hydraulic behaviors typical of these safety systems, including the primary depressurization due to natural circulation cooling, a nonuniform flow behavior among SG U-tubes, an accumulation of the non-condensable gas originally contained in the injected water, liquid holdup in U-tubes due to the countercurrent flow limiting, and long-term passive core cooling with the GDIS injection. From the assessment of the RELAP5/MOD3 code using the present data, it was found that the inability of the code to predict the U-tube nonuniform flow behavior resulted in overprediction of the primary depressurization rate at a pressure less than 1 MPa, and exaggerated oscillation of the natural circulation flow rate in the primary loop.     

Prediction of counter-current flow limitation at hot leg pipe during a small-break LOCA

The possibility of hot leg flooding during reflux condensation cooling after a small-break loss-of-coolant accident in a nuclear power plant is evaluated. The vapor and liquid velocities in hot leg and steam generator tubes are calculated during reflux condensation cooling with the accident scenarios of three typical break sizes, 0.13, 1.02 and 10.19% cold leg break. The effect of initial water level to counter-current flow limitation is taken into account. It is predicted that the hot leg flooding is precluded when all steam generators are available for heat removal. It is also shown that both hot leg flooding and SG flooding are possible under the operation of one steam generator. Therefore, it can be said that the occurrence of hot leg flooding under reflux condensation cooling is possible when the number of steam generators available for heat removal is limited.     

自然循环蒸汽发生器并联倒U型管流量分配计算

针对自然循环工况下蒸汽发生器部分倒U型管内存在倒流现象,通过对倒U型管内流动传热特性进行分析,获得了倒流发生的判断依据,从而编制了流量分配计算程序。采用该程序对某型蒸汽发生器并联倒U型管流量分配进行了计算,通过将结果与实验值进行对比分析,对程序可信度进行了验证,并采用该程序对蒸汽发生器并联倒U型管主要热工参数随进出口压降变化情况进行了计算分析。结果表明,倒流现象发生在短管内,倒流的发生使得蒸汽发生器一次侧净流量和单位时间输热呈阶梯下降,对反应堆安全产生较大的影响。     

PWR Cold-Leg Small-Break LOCA with Total HPI Failure

Cold-leg small-break loss-of-coolant accident (LOCA) tests were performed at the ROSA-IV Large Scale Test Facility (LSTF), a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). The tests were conducted for break areas ranging 0.5–10% of the scaled cold leg area, and simulated hypothetical total failure of the high pressure injection (HPI) system. One of the tests, conducted with 1% break area, included an intentional depressurization of the primary system that was initiated after the onset of core dryout. A simple prediction model is proposed for prediction of times of major events, namely, loop seal clearing, core dryout, accumulator (ACC) injection and actuation of low pressure injection (LPI) system. Test data and model calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of approximately 5% or more, and might be insufficient for intermediate break areas to maintain adequate core cooling. It is also shown that there might be possibility of core dryout after ACC injection and before LPI injection for break areas less than approximately 2.5%.     

蒸汽发生器U型管低含气率两相倒流现象研究

针对立式倒U型管自然循环蒸汽发生器传热管内的两相倒流现象,基于均相流模型,建立了U型管内低含气率两相流动传热理论模型,给出了U型管的进出口压降-质量流量曲线,分析了U型管内出现两相倒流现象的机理,研究了二次侧流体温度和入口含气率对倒流现象的影响规律,并与单相倒流进行了对比。利用RELAP5/MOD 3.3程序对相同条件下的倒流问题进行了计算。研究表明,提高蒸汽发生器二次侧工作压力可减少倒流,两相流入口含气率越高,倒流越易发生,两相流较单相流在U型管内更易倒流。     

自然循环条件下蒸汽发生器U型传热管倒流分布特性的实验研究

在自然循环工况下蒸汽发生器一次侧入口流量为0.4~0.7 kg/s的参数范围内，开展了蒸汽发生器U型传热管倒流特性实验。针对9种不同长度的U型传热管，分别设置9个倒流监测点，获得了倒流在不同长度U型管中的分布特性。基于传热管压降实测数据和守恒原理，获得了蒸汽发生器一次侧的倒流总流量以及倒流U型管的数目。结果表明，在本实验参数范围内，约有61%的U型管发生倒流，使传热管正向流通面积减小为原来的39%。倒流同时导致正流流量增加60%，与不发生倒流的情况相比，U型管平均流速增大4.2倍。      

Effects of fluid properties on CCFL characteristics at a vertical pipe lower end

The purpose of this study is to derive a counter-current flow limitation (CCFL) correlation and evaluate its uncertainty for steam generator (SG) U-tubes in a pressurized water reactor (PWR). Experiments were conducted to evaluate effects of the liquid viscosity on CCFL characteristics using air–40 wt% or air–60 wt% glycerol water solution and saturated steam–water at atmospheric pressure with vertical pipes simulating the lower part of the SG U-tubes. The steam–water experiments confirmed that CCFL characteristics could be expressed in terms of the Wallis parameters (J_G* and J_L*) for the pipe diameters of D = 14, 20, and 27 mm. A CCFL correlation was derived using the ratio μ_G/μ_L of the viscosities of the gas and liquid phases, μ_G and μ_L, as a correction term representing effects of fluid properties, where J_G^*1/2(μ_G/μ_L)^?0.07 was expressed by a cubic function of J_L^*1/2(μ_G/μ_L)^0.1. In the correlation, the constant C indicating the value of J_G^*1/2(μ_G/μ_L)^?0.07 at J_L* = 0 was (1.04 ± 0.05), and this uncertainty of ±0.05 would cover most of the previous experimental data including the ROSA-IV/LSTF data at 1, 3, and 7 MPa.     

自然循环蒸汽发生器倒U型管内倒流现象影响因素研究

在某些自然循环工况下,蒸汽发生器部分倒U型管内存在倒流现象。基于一维Oberbeck-Boussinesq方程,建立了蒸汽发生器并联倒U型管内单相水流动传热模型,并以两种尺寸的蒸汽发生器为例进行了计算。计算结果表明,小型蒸汽发生器内短管易发生倒流,大型蒸汽发生器内长管易发生倒流;蒸汽发生器进口水温对倒流现象的发生具有重要的影响。     

Heat Transfer in a Fluid with Internal Heat Generation Flowing through a Vertical Tube

This paper deals with heat transfer in a fluid, with uniformly distributed internal heat source, flowing upwards through a vertical tube. Measurements were made of the temperature distribution in both laminar and turbulent flow, and both with and without heat transfer at wall. Heat generation within the fluid was brought about by passing an electrical current through the working fluid, which was an aqueous solution of sodium chloride. The experimental results were compared with analytical calculations.

Free convection, which occurs and is superimposed on the movement by forced convection, flattens the fluid temperature distribution in laminar flow through thermally insulated vertical tube. In turbulent flow with heat transfer to wall, the temperature distribution near the wall is affected considerably by the outgoing heat flux.     

Analysis of System Thermal Hydraulic Responses for Passive Safety Injection Experiment at ROSA-IV/Large Scale Test Facility

An experiment was conducted at the ROSA-IV/Large Scale Test Facility (LSTF) on the performance of a gravity-driven emergency core coolant (ECC) injection system attached to a pressurized water reactor (PWR). Such a gravity-driven injection system, though not used in the current-generation PWRs, is proposed for future reactor designs. The experiment was performed to identify key phenomena peculiar to the operation of a gravity injection system and to provide data base for code assessment against such phenomena. The simulated injection system consisted of a tank which was initially filled with cold water of the same pressure as the primary system. The tank was connected at its top and bottom, respectively, to the cold leg and the vessel downcomer. The injection into the downcomer was driven primarily by the static head difference between the cold water in the tank and the hot water in the pressure balance line (PBL) connecting the cold leg to the tank top. The injection flow was oscillatory after the flow through the PBL became two-phase flow. The experiment was post-test analyzed using a JAERI modified version of the RELAP5/MOD2 code. The code calculation simulated reasonably well the system responses observed in the experiment, and suggested that the oscillations in the injection flow was caused by oscillatory liquid holdup in the PBL connecting the cold leg to tank top.     

U型管内蒸汽冷凝回流压降特性的研究

进行了单根 U型管内蒸汽冷凝回流实验。 U型管的内径为 20mm,总高度为 4.1m和 7.0m两种。在系统压力 0.1～ 6.0MPa、蒸汽质量流速 4～ 45kg/m2· s、二次侧进口冷却水温度 20～ 196℃的范围内,研究了 U型管内蒸汽冷凝回流的流动及其压降特性。     

PWR pressurizer modeling

A series of experiments have been performed which help to provide fundamental understanding of the phenomena which are important to the analysis of a PWR pressurizer. The transients considered include insurges to a partially-full tank, outsurges, insurges to a tank with hot walls, empty tank insurges, and combined insurges and outsurges. The experiments include the effects of noncondensable gases, and free surface heat transfer. These experiments provide a data base from which recommendations are made for calculating such phenomena, as: (i) stratification of the hot water and incoming cold water, (ii) wall condensation, (iii) flashing, (iv) rainout, (v) suppression of flashing, (vi) wall conduction, (vii) the effect of noncondensable gases on wall heat transfer, and (viii) free surface heat transfer. From these experiments a general model of a PWR pressurizer has been developed. It will predict the pressure-time behavior of a PWR pressurizer during a variety of transients. The model has been benchmarked against the low pressure experiments of this study and a single full scale pressurizer transient experiment. The most significant finding is the pool in the pressurizer remains stratified during an insurge transient so that practically no heat transfer to the cold insurge liquid occurs. Wall heat transfer can be significant for insurge transients, however.     

一体化压水堆蒸汽发生器的热工水力瞬态特性分析

一体化压水堆的设计是将蒸汽发生器及稳压器等一回路所有部件都放入压力容器内，以提高安全性，采用可以精确模拟直汉蒸汽发生器二次侧水的饱和点，蒸干点位置等重要参数随时间变化的可移动边界并分法，选用适合各中换热工况的一整套换热关系式，建立了可以模拟一体化压水堆直流蒸汽发生器的稳态及瞬态热工不特性的物理及数学模型，并编制了计算程序，经对Ｂａｂｃｏｃｋ和Ｗｉｌｃｏｘ公司１９管直流蒸汽发生器实验装置进行了计算有     

Research on the effect of Reynolds correlation in natural convection film condensation

Film condensation is a vital phenomenon in the nuclear engineering applications,such as the gas-steam pressurizer design,and heat removing on containment in the case of postulated accident.Reynolds number in film condensation can be calculated from either the mass relation or the energy relation,but few researches have distinguished the difference between them at present.This paper studies the effect of Reynolds correlation in the natural convection film condensation on the outer tube.The general forms of the heat transfer coefficient correlation of film condensation are developed in different flow regimes.By simultaneously solving a set of the heat transfer coefficient correlations with Remass and Reenergy,the general expressions for Remass and Reenergy and the relation between the corresponding heat transfer coefficients are obtained.In the laminar and wavefree flow regime,Remass and Reenergy are equivalent,while in the laminar and wavy flow regime,Remass is much smaller than Reenergy,and the deviation of the corresponding average heat transfer coefficients is about 30％ at the maximum.In the turbulent flow regime,the relation of Remass and Reenergy is greatly influenced by Prandtl number.The relative deviation of their average heat transfer coefficients is the nonlinear function of Reynolds number and Prandtl number.Compared with experimental results,the heat transfer coefficient calculated from Reenergy is more accurate.     

蒸汽发生器传热管一、二次侧耦合换热及管外过冷沸腾数值研究

采用两流体欧拉数学模型,结合气相和液相之间的界面传热、传质和动量交换封闭模型以及RPI壁面沸腾模型,利用ANSYS CFX 12.0对蒸汽发生器局部传热管束二次侧的过冷沸腾进行数值研究。数值研究结果与单管内过冷沸腾实验数据对比验证符合良好。结果表明,采用壁面沸腾模型能准确预测沸腾起始点的位置,同时梅花孔板的存在对二次侧流动换热特性影响显著。     

非凝性气体对汽-气稳压器稳压的影响研究

通过分析相间的传热传质过程以及非凝性气体存在时壁面蒸汽冷凝过程,建立了汽 气稳压器模型,研究了非凝性气体对稳压过程的影响,描述了稳压器的稳压特性,并将模型计算结果与MIT稳压器实验数据进行了对比。结果表明：当不含非凝性气体时,计算精度高,相对偏差在0.8%内,压力峰值为0.647 MPa;当非凝性气体含量从0增至20%时,计算精度相对减小,最高相对偏差为15.4%;压力峰值从0.647 MPa增至1.02 MPa。研究表明非凝性气体对稳压器稳压过程具有重要影响作用,随着非凝性气体的种类和含量的变化,稳压器内稳压过程发生显著变化。     

Analysis of the OECD/NEA ROSA Project experiment simulating a PWR small break LOCA with high-power natural circulation

OECD/NEA ROSA Project experiment with the large scale test facility (LSTF) in JAEA was conducted simulating a PWR 1% cold leg small break LOCA with an assumption of high-power natural circulation due to failure of scram and total failure of high pressure injection system. The core power curve for the LSTF experiment was obtained by PWR LOCA analysis using JAEA-developed coupled three-dimensional kinetics and thermal-hydraulics code SKETCH-INS/TRAC-PF1 with detailed core model. A post-test analysis was performed against the obtained data by using JAEA-modified RELAP5/MOD3.2.1.2 code to validate the code predictability. The JAEA-modified RELAP5/MOD3.2.1.2 code was used by incorporating a break model that employs maximum bounding flow theory with a discharge coefficient of 0.61 for two-phase break flow. In the experiment, flow in hot legs became supercritical during two-phase natural circulation, causing the hot leg liquid level to be quite low. Liquid accumulation in steam generator U-tube upflow-side took place during reflux condensation mode due to high vapor velocity. The RELAP5 code predicted reasonably well the overall thermal-hydraulic phenomena observed in the experiment. The code, however, overpredicted the break flow rate especially during two-phase flow discharge period probably because of the failure in the correct simulation of the cold leg liquid level due to late decrease in the primary loop flow rate.