倾斜并联内螺纹管汽-水两相流密度波型脉动试验研究

在系统压力p=3～10MPa,质量流速G=300～600kg/s,进口过冷度Δtsub=30～90℃,内壁热负荷q=0～190kW/m2的工况范围内,采用试验段长度与内径之比(L/d)大于600、倾角为19.5o的φ38.1×7.5mm6头内螺纹管,研究了压力、质量流速、进口过冷度以及两管热负荷不均匀对高压汽-水两相流密度波脉动的影响。结果表明,随压力增加,系统稳定性增加;随质量流速增加,临界热负荷增加,而临界干度下降;进口过冷度对密度波脉动呈现单值性影响,随进口过冷度下降,临界热负荷降低;在其他条件相同的情况下,并联管不对称加热时的临界热负荷较对称加热时的临界热负荷更高。     

含不凝性气体蒸汽浸没射流冷凝压力振荡实验研究

实验研究了不凝性气体（空气）含量、水温和蒸汽质量流速对蒸汽浸没射流冷凝压力振荡特性的影响,实验工况横跨冷凝振荡（CO）区和稳定冷凝（SC）区。结果表明：对于纯蒸汽射流,压力振荡主频随水温的升高而降低,振荡强度随水温的升高而升高;在CO区,振荡主频和振荡强度均随蒸汽质量流速的升高而升高;在SC区,振荡主频随蒸汽质量流速的升高而降低,振荡强度基本上不随蒸汽质量流速的变化而发生改变;对于含空气射流,随空气质量分数的增加,振荡主频总体呈下降趋势,振荡强度先迅速下降后小幅上升,在空气质量分数为0.05～0.1区域内振荡主频和振荡强度均存在极小值。     

非对称工况下并联通道流动不稳定性研究

针对并联矩形通道,基于集总参数法建立了并联通道流动不稳定性分析模型,并基于此模型分析了非对称工况对系统稳定性的影响。分析结果表明,非对称工况对并联通道流动不稳定性有显著影响,在保持平均节流系数恒定的情况下,非对称节流的影响随压力增大而降低,随质量流速增大而增大;界限功率随加热不对称度增加呈先上升后下降趋势,且非对称加热的影响随压力增大而增大,随入口过冷度和质量流速增大而减小。     

低质量流速下不同尺寸垂直并联内螺纹管压力降不稳定试验研究

在高压汽水试验回路上进行低质量流速下两相流压力降不稳定对比试验研究,分析进口压力、质量流速、进口过冷度对两种垂直并联内螺纹管两相流不稳定的影响.结果表明,两种内螺纹管发生压力降型脉动的界限干度范围和脉动周期范围接近;随进口压力、质量流速的增加,发生压力降型脉动的临界热负荷增加;随着进口压力的增加,发生压力降型脉动的界限...     

高温气冷堆He载气中H2O和CO2吸附净化的实验研究

通过对高温气冷堆Ｈｅ载气中的Ｈ２Ｏ和ＣＯ２在５Ａ分子筛固定床上吸附净化的实验研究，得到了吸附穿透曲线，获得了５Ａ分子筛床对Ｈ２Ｏ和ＣＯ２的动态吸附规律。实验考察了吸附温度、工作压力、杂质浓度、流速及床层高度等因素对Ｈ２Ｏ和ＣＯ２单吸附及共吸附的吸附容量及转效时间的影响，获得了最佳运行参数。实验研究结果表明：净化后Ｈｅ气中Ｈ２Ｏ和ＣＯ２的质量分数低于１０－５，满足了净化系统的要求，为高温气冷堆中５Ａ分子筛固定床装置提供了可靠的设计数据。     

蒸汽浸没射流压力振荡特性实验研究

为指导Ⅰ型鼓泡器排放管的结构设计,对常压工况下不同蒸汽质量流速（150～500 kg/(m²·s)）、池水过冷度（18～68 ℃）、开孔直径（10、16 mm）和孔数（单孔、双孔）蒸汽浸没射流的压力振荡特性开展实验研究。结果表明：不同开孔结构下的振荡强度均随蒸汽质量流速的升高或池水过冷度的降低而先增大后减小;当孔径增大或孔数增多时,达到稳定冷凝所需的最小蒸汽质量流速降低,流型转变导致振荡强度减小。不同开孔结构下的振荡主频均主要受池水过冷度和开孔直径的影响,且随过冷度的降低或孔径的增大而减小。孔数变化对主频的影响是多重的,但总体效果弱于孔径变化。通过对主频实验值进行拟合,获得了由池水过冷度和排放孔径表征的半经验关系式,相对误差在±20%以内。     

双矩形通道密度波不稳定性研究

针对窄间隙矩形通道的密度波不稳定性进行了试验研究,试验采用了断面尺寸为25 mm×2 mm,加热长度为1 000 mm的双扁形矩形通道组成试验段.试验结果发现增加质量流速、入口过冷度、压力,均能增加稳定流动的范围.脉动周期随质量流速的增加而变短,随入口过冷度的增加而增加.压力对脉动周期的影响较小.以无量纲过冷度Nsub和相变数Npch比较不同长度试验段的结果,发现平行矩形通道的结果和平行直圆管的结果基本重合,长度和流道断面形状对流动不稳定性的影响较小.     

蒸汽射流凝结压力振荡幅值研究

为了得到蒸汽在过冷水中浸没射流凝结引起的压力振荡特性,针对不同的蒸汽质量流率和过冷水温度对压力振荡的影响进行实验研究。通过高频压力传感器测量得到了不同测点位置的压力振荡幅值,并结合蒸汽射流凝结形态分析蒸汽质量流率和过冷水温度对压力振荡幅值的影响规律。结果表明,蒸汽射流凝结在凝结振荡区比较剧烈,在稳定射流区相对缓和,从凝结振荡向稳定射流的形态转变导致了压力振荡幅值随蒸汽质量流率出现先减小后增大的趋势;振荡幅值随水温的升高不断增大,特别是当水温大于40℃时,由于过冷度减小,射流凝结形态变得发散,导致凝结振荡幅值大幅增加。     

近临界压力区垂直内螺纹管临界热流密度实验研究

在近临界压力区,对垂直上升内螺纹管流动沸腾的偏离泡核沸腾(DNB)型临界热流密度(CHF)现象进行了实验研究。试验段采用ф35 mm×5.67 mm六头内螺纹管。实验参数范围为:压力18~21 MPa,质量流速500~1 000kg/(m~2·s),进口过冷度3~5℃,内壁热负荷40~960kW/m~2。实验得到了不同工况下的内壁温度和传热系数分布特性,分析了流动参数对内螺纹管中DNB型CHF的影响,并根据实验数据拟合出两相区的传热关联式与临界热流密度(qCHF)预测关联式。内螺纹管的qCHF实验数据被用于与光管的qCHF预测值进行对比,发现内螺纹管具有一定的CHF强化作用,但当压力越靠近临界压力时这种作用会被抑制甚至消失。实验结果表明:在近临界压力下,内螺纹管会在低干度区甚至过冷区发生DNB现象,压力的增大和质量流速的减小均会使DNB提前发生。qCHF随压力的减小和质量流速的增大而增大。在特定工况下,试验段不同截面会分别发生偏离泡核沸腾与蒸干。     

热流密度对汽水两相流压力波动特性的影响

实验研究了螺旋管内绝热和沸腾汽液两相流动时压力波动特性。分析了热流密度对压力波动的影响规律。结果表明沸腾时波动的均方根和分形参数（分维数、关联维数和Kolmogorov熵）与绝热两相流明显不同,加热热流密度的大小影响各特征参数的大小及其随干度的变化规律。未考虑热流密度影响的流型图或流型转变准则只能在小热流密度时使用。     

An experimental study on the critical heat flux for low flow of water in a non-uniformly heated vertical rod bundle over a wide range of pressure conditions

An experimental study of the critical heat flux (CHF) has been performed for a water flow in a non-uniformly heated vertical 3 × 3 rod bundle under low flow and a wide range of pressure conditions. The experiment was especially focused on the parametric trends of the CHF and the applicability of the conventional CHF correlations to a return-to-power conditions of a main steam line break accident whose conditions might be a low mass flux, intermediate pressure, and a high inlet subcooling. The effects of the mass flux and pressure on the CHF are relatively large and complicated in the low pressure conditions. At a high mass flux or a low critical quality, the local heat flux at the CHF location sharply decreases with an increasing local critical quality. However, at a low mass flux or a high critical quality, the local heat flux at the CHF location shows a nearly constant value regardless of the increase of the critical quality. The CHF data at the very low mass flux conditions are correlated well by the churn-to-annular flow transition criterion or the flow reversal phenomena. Several conventional CHF correlations predict the present return-to-power CHF data with reasonable accuracies. However, the prediction capabilities become worse in a very low mass flux of below about 100 kg/(m² s).     

竖直圆管内向上流动的干涸实验研究

在直径为8.2 mm的竖直向上均匀加热圆管上进行了干涸型临界热流密度实验研究,加热长度2.4 m,压力3.2～19.7 MPa,质量流速963～2 707 kg/（m²•s）,进口欠热度34～213 ℃,出口含汽率0.11～0.78。研究发现：临界热流密度随进口欠热度、质量流速的增加而线性增加,随出口含汽率的增加而迅速减小。通过对临界气液两相参数的分析发现,本实验参数范围内,蒸汽速度是导致干涸的主要原因。当达到临界蒸汽速度时,近壁面液体消失触发临界。随压力的增加,表面张力逐渐减小,液膜更易被撕裂,因而临界蒸汽速度随压力的增加而减小。从高压实验数据出发得到临界蒸汽速度U_cr,参考Steen和Wallis推荐的夹带开始速度U₀,建立了预测临界蒸汽速度的模型：U_cr=25U₀+4。利用低压实验数据对预测模型进行了验证,符合较好。     

内螺纹管临界热流密度预测与数值计算

基于壁面汽泡壅塞理论,针对近临界压力区两相流动沸腾的偏离泡核沸腾(DNB)现象,对垂直上升内螺纹管的DNB型临界热流密度(CHF)进行了数值计算研究。以内螺纹管为分析对象改进已有的汽泡壅塞模型,计算了汽泡层区与主流区的极限传递质量流量、湍流速度分布、汽泡层区临界截面含气率等本构关系,汽泡脱离直径的计算考虑了汽泡接触角的影响。本文模型还根据大量CHF实验数据拟合得到了新的α_b关联式。最后,基于Fortran语言编制了CHF的理论预测数值模型程序,研究分析了压力、质量流速、热平衡干度及进口欠焓对CHF的影响,并根据CHF查表值对本文模型进行评估,同时将实验得到的内螺纹管CHF数据与采用Bowring模型、Katto模型、Shah模型和本文模型计算的CHF进行比较,发现本文模型的误差最小,与实验值吻合结果较好,说明本文模型能较好地对垂直上升内螺纹管DNB型CHF进行预测。     

Experimental research on dryout point of flow boiling in narrow annular channels

1 Introduction There are basically two classes of critical heat flux (CHF) situations: departure from nucleate boiling (DNB) and dryout (DO) [1]. DO is also sometimes known as burnout or departure from forced convective boiling in vapor-continuous flow. From the point of view of engineering, the CHF caused by the DO mechanism is of particular importance since boiling annular flow is one of the most common flow patterns in gas–liquid two-phase flow and occurs in a wide range of vapor qua…     

Effects of mechanical vibration on critical heat flux in vertical annulus tube

The effects of mechanical vibrations on critical heat flux (CHF) are examined in this study at atmospheric pressure in vertical annulus tube under electrically heated condition. Vibration of heating rod was increased as flow regime changed from subcooled region to bubbly region. CHF was increased by mechanical vibration up to 16.4%. Vibration amplitude was one of the effective parameters on CHF enhancement. It seems to come from turbulence increasing and increment of deposition of droplet from the liquid film by vibration. Vibration is an effective method for heat transfer enhancement as well as CHF enhancement.     

An investigation on heat transfer characteristics of different pressure steam-water in vertical upward tube

Within the range of pressure from 9 to 30 MPa, mass velocity from 600 to 1200 kg/(m² s), and heat flux at inner wall from 200 to 600 kW/m², experiments have been performed to investigate the heat transfer characteristics of steam-water two-phase flow in vertical upward tube. The outer diameter of the tube is 32 mm, and the wall thickness is 3 mm. Based on results, it was found that Dryout is the main mechanism of the heat transfer deterioration in the sub-critical pressure region. Near the critical pressure, when the heat transfer deterioration occurs, the steam quality of water is lower than that in the sub-critical pressure region, so that DNB is the main mechanism in this pressure region. At supercritical pressure, the heat transfer performance in circular channel is improved and enhanced. Heat transfer deterioration phenomenon is observed when the fluid bulk temperature approaches to the pseudo-critical value. Nusselt correlation of the forced-convection heat transfer in supercritical pressure region has been provided, which can be used to predict heat transfer coefficient of the vertical upward flow in tube.     

Investigation on the characteristic of CHF in various flow pattern regimes based on look-up table data

The critical heat flux (CHF) is one of the important phenomena limiting the maximum rate of heat transfer and hence power rating of nuclear reactors. The thermal hydraulic phenomena like pressure drop, heat transfer, stability, etc. depends upon the flow pattern in the system. The CHF phenomenon is also closely related to the two-phase flow patterns. It is important to investigate the dependence of CHF on the flow pattern regimes to understand the underlying mechanisms. The present investigation reveals that CHF generally increases with mass flux in the churn/slug region. However, in the annular region the CHF decreases with increase in mass flux. Considering the dependency of the CHF trend on the flow pattern regime, it will be useful to develop CHF models, which are specific to the flow pattern regime. The data of CHF look-up table has been considered in this investigation since this approach is one of the most reliable methods for the prediction of CHF and is being used in several best-estimate thermal-hydraulic system codes, such as RELAP5, CATHARE and CATHENA. The pressure, mass flux and quality have been considered as important thermal hydraulic parameters to characterize the flow pattern during CHF under various operating condition.     

Critical Heat Flux in a Heater Rod Bundle Cooled by R-134a Fluid near the Critical Pressure

In the development of supercritical pressure water cooled reactors, it is important to understand the characteristics of a heat transfer near the thermodynamic critical point. An experimental study on the critical heat flux near the critical pressure has been performed with a 5 × 5 square array heater rod bundle cooled by R-134a fluid (P _c = 4:059MPa, T _c = 101°C). The critical power has been accurately measured up to the reduced pressure of 0.99 (4.03 MPa). The critical power decreases sharply at a pressure of about 3.8–3.9 MPa as the pressure approaches the critical pressure. For the low mass fluxes of 50 to 250kg/m², a sharp decrease in the critical power is not observed near the critical pressure. The CHF phenomenon near the critical pressure no longer leads to an inordinate increase in the heated wall temperature such as the case of DNB at normal pressure conditions. In the pressure region close to the critical pressure, there is a threshold pressure at which the CHF phenomenon disappears. When the pressure exceeds the threshold pressure, the wall temperature increases monotonously without a CHF occurrence according to the power level applied to the heater rods. The threshold pressure moves toward the lower pressure region gradually with an increasing mass flux.     

Prediction of critical heat flux in vertical pipe flow

A previously developed semi-empirical model for adiabatic two-phase annular flow is extended to predict the critical heat flux (CHF) in a vertical pipe. The model exhibits a sharply declining curve of CHF versus steam quality (X) at low X, and is relatively independent of the heat flux distribution. In this region, vaporization of the liquid film controls. At high X, net deposition upon the liquid film becomes important and CHF versus X flattens considerably. In this zone, CHF is dependent upon the heat flux distribution. Model predictions are compared to test data and an empirical correlation. The agreement is generally good if one employs previously reported mass transfer coefficients.