R134a卧式螺旋管内流动沸腾CHF特性研究

为探索低潜热工质在卧式螺旋管内流动沸腾的临界热流密度(CHF)特性,采取大电流直接对实验段通电,利用不锈钢电阻加热的方式,在出口压力p=0.40~1.05 MPa,质量流速G=51~257 kg/m2s,入口热平衡干度Xi=-0.18~0.43的条件下开展了R134a在卧式螺旋管内流动沸腾的CHF特性研究。实验用螺旋管内径7.6 mm,螺旋径300 mm,节距40 mm,有效加热长度7.07 m。重点分析了实验段壁温沿管长和管截面周向的变化规律,以及压力、流量、干度等参数对CHF值的影响,并把实验数据与Bowring和Shah关联式的计算值进行了验证比较,结果发现这2种关联式在该实验条件下均不适用。     

竖直矩形窄缝通道内过冷沸腾传热特性的实验研究

以去离子水为工质,在进口压力为0.1～0.3 MPa、质量流速为200～1400 kg·m^-2·s^-1、热流密度为20～320k W·m^-2的参数范围内,对截面参数为50mm×2mm的竖直矩形窄缝通道展开了传热实验研究。实验获得通道内部工质由单相状态到过冷沸腾状态的传热过程曲线,将过冷沸腾段实验值与8个经验公式提供的预测值进行了对比与分析,采用相似原理以及回归分析法,建立了适用于竖直矩形窄缝通道的过冷沸腾准则关系式。研究结果表明,在竖直矩形窄缝通道内,热流密度对过冷沸腾传热具有主导作用;对于本实验的窄缝通道,Bertsch传热公式对于过冷沸腾段的预测效果相较于其他公式更好,本研究所建立的准则关系式与实验数据符合良好。因此,本研究建立的公式能够用于竖直矩形窄缝通道过冷沸腾传热系数的预测。     

卧式螺旋管内流动沸腾传热研究

在较宽广的参数范围内对卧式螺旋管内水／水蒸汽两相流沸腾传热特性进行了详细的试验研究，螺旋管内径ｄ＝１１ｍｍ，曲率直径比Ｄ／ｄ为２３．２７。试验参数范围：压力０．５－３．０ＭＰａ：质量流速２００－２５００ｋｇ／ｍ２·ｓ；热负荷２３０－５００ｋＷ／ｍ２：出口干度０－０．８６。得到了沸腾传热系数的实验关联式，并深入分析了局部传热与壁温特性，沿流动方向，管截面平均传热系数在螺旋管的上升段最大，而在下降段最小；沿管截面内表面圆周方向上，局部传热系数在外侧最大，在内侧最小，在顶部和底部处的值居中。得到了管截面局部传热系数的分布实验关联式。同一管截面外表面圆周方向上，外侧壁温最低而内侧最高。     

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。     

螺旋管内汽液两相流密度波型脉动流动传热试验研究

详细介绍了在沸腾通道内部发生汽液两相流水动力不稳定性而出现周期性密度波型脉动时,脉动流动过程中瞬态和时均传热系数的实验研究结果。实验在以水为工质、以螺旋管作沸腾蒸发试验段的中低压闭式循环系统上进行,试验参数范围为：压力p=05～35 MPa,质量流速G=200～2 100 kg/(m2·s),工质进口过冷度ΔTsub=20～90 ℃,试验段壁面热负荷qw=0～540 kW/m2,密度波脉动的周期为T=125～14 s,且主要集中在4～10 s范围内。对密度波脉动过程中瞬态及时均传热系数和其它主要参数的基本特征与变化规律作了分析和描述,提出了表征密度波脉动传热的新的特征准则数和传热系数计算式。     

在垂直环形窄缝流道中的沸腾传热特性研究

为了弄清在窄缝环形流道中气泡的形成、聚合和变形的特性 ,以及气泡在聚合变形之后对传热特性的影响 ,在常压下用蒸馏水对窄缝间隙为 0 75mm的垂直环形流道 ,进行了可视化的流动沸腾传热实验研究 ;实验段的有效加热长度为 90 0mm ,其加热方式为单面内侧加热 ,实验的流量变化范围为 1 667× 1 0 - 5m3/s至 5 833× 1 0 - 5m3/s。实验得到了在不同质量流密度和热流密度下窄缝流道中的沸腾传热系数随干度变化的分布。通过与常规流道中的沸腾传热系数的比较 ,得到了在窄缝环形流道中沸腾传热系数比常规流道中的沸腾传热系数约高 1 5 %的结论。另外通过用高速摄像机对可视化的垂直环形流道中的流型进行的拍摄研究 ,分清了存在在窄缝环形流道中的四种流型     

单棒通道流动振荡诱发沸腾临界可视化实验研究

为深入分析沸腾两相流动振荡诱发沸腾临界的影响特性,本文以去离子水为工质,横截面19 mm×19 mm、中心为外径9.5 mm的单棒通道为研究对象,通过在不同热工参数下开展沸腾两相流动特性可视化实验研究,结合汽泡行为和汽-液界面特性,分析流动振荡诱发沸腾临界的影响特性。研究结果表明,低压力、低质量流速和低入口过冷度下,极易出现流动振荡,并导致沸腾临界提前发生,此时的临界热流密度与稳定工况下相比明显偏低;随着壁面热流密度不断增加,流道中两相流型先后出现泡状流、弹状流、合并弹状流、搅混流、剧烈搅混流、不稳定环状流;当流动出现剧烈振荡时,流道存在回流;发生沸腾临界时流道压降波动最大,对应的流型为不稳定环状流。因此,单棒通道内流动振荡可能会导致沸腾临界提前发生。      

环形通道内液态金属钠沸腾两相传热特性实验研究

对环形通道内液态金属钠沸腾两相流动特性进行了实验研究。实验中,系统压力为3.6~110.0kPa,热流密度为11~600kW·m~(-2),流速为0.02~0.45m·s~(-1)。实验结果表明,液态金属钠沸腾传热系数与壁面热流密度和系统压力有强烈关系,而与入口过冷度和质量流速无关。在本文实验数据基础上,拟合得到了计算液态金属钠沸腾两相传热系数的关系式,通过与各组实验数据间的比较,证明本文关系式适用于计算环形通道内液态金属钠沸腾两相传热系数。     

环形通道内钠流动起始沸腾壁面过热度实验研究

对环形通道内金属钠起始沸腾壁面过热度进行实验研究。实验段长800 mm,环形通道外径10 mm,内径6 mm。电加热元件最高热流密度为846 kW/m²,进口过冷度为63.1～287.8 ℃,质量流量为7.2～122.0 kg/h,系统压力为0.85～28.79 kPa。实验结果表明,起始沸腾壁面过热度随热流密度和进口过冷度的增加而升高,随质量流量和系统压力的增加而降低。拟合得到了关于起始沸腾壁面过热度的半经验关系式,关系式计算结果与实验数据符合良好。     

基于漂移流模型的液态金属快堆螺旋管蒸汽发生器热工水力特性分析

在液态金属快堆螺旋管蒸汽发生器中,存在一个普遍问题,其一次侧的进、出口温差大幅升高,二次侧出口蒸汽过热度显著增大,这给其设计及运行带来了挑战。基于离散网格法建立了液态金属快堆螺旋管蒸汽发生器热工水力分析模型。模型对整个一、二次侧回路进行网格划分,采用漂移流模型计算二次侧水-水蒸汽的流动与传热,并在一次侧计算中采用液态金属物性与流动传热关联式;采用内节点法对壁面划分网格,考虑两侧流体与管壁间的对流换热以及壁面导热。基于实验数据验证模型可靠性。以铅铋快堆为例,研究不同入口条件下蒸汽发生器的热工水力特性。研究发现一、二次侧之间的壁面热流密度沿程分布极为不均匀,且热流密度峰值极高。算例中壁面热流密度最大值达到1361 kW/m²,最大值与最小值间相差数十倍到数百倍。随着一次侧入口铅铋温度以及铅铋流速的增加,二次侧过冷水区及两相区长度明显缩短,过热蒸汽区长度明显增大;同时,壁面热流密度峰值向螺旋管入口方向移动,二次侧工质压降明显增大。     

垂直管内汽水两相下降流动和上升流动时的沸腾传热特性

对垂直管内汽水两相下降流动和上升流动时的沸腾传热特性进行了理论和试验研究.推导得到了两种流动情况下的汽泡脱离半径,分析讨论了液膜厚度及流速,从而对试验结果作出了理论解释.研究结果表明,在临界压力以下,上升流的传热恶化比下降流严重;在超临界压力下,上升流传热特性好于下降流.     

Combinated Flow Pattern Map for Cocurrent and Countercurrent Air-Water Flows in Vertical Tube

Flow patterns for cocurrent and countercurrent air-water flows in vertical tubes (40 and 80mm I.D.) at volumetric flux densities of air and water in the ranges ?115–158 and ?100–102 cm/s were observed. A flow pattern map presenting the entire data of the observed flow patterns, i.e. bubbly, slug and annular flow for each mode of flow operation (upflow, countercurrent flow and downflow) were presented on the j_l vs. j_g plane. The flow pattern maps showed significant difference of flow pattern transition boundaries with upflow, countercurrent flow and downflow. Flow pattern transition curves were smoothly continuous with the change of the direction of water flow, on the other hand the change of flow direction of air showed complicated effect on flow pattern transition near zero j_g . Comparison of the present flow pattern data with the reported general flow pattern maps for upflow showed that the correlation of Taitel et al. for bubble-slug flow transition is applicable to both cocurrent and countercurrent air-water flows.     

Radiation Anneal Hardening of Neutron Irradiated Niobium

Experiments were carried out with a vertical rectangular channel simulating a sub-channel of the upgraded JRR-3 fuel element, in order to investigate the validity and the error of the correlations predicting the superheat at the onset of nucleate boiling. These correlations were used in the core thermal-hydraulic design of the upgraded JRR-3. As the results, the following were made clear: (1) The existing Bergles-Rohsenow correlation gives a good prediction for the relationship of heat flux vs. superheat at the onset of nucleate boiling, with the error of about 1 K against the lower limits of the measured superheat. (2) There are no significant differences in the characteristics of the relationship of heat flux vs. superheat at the onset of nucleate boiling between upflow and downflow. (3) There are no significant differences in the histories of relationship of heat flux vs. superheat from the forced convection single-phase flow to the subcooled boiling between increasing heat flux and decreasing heat flux, with little overshoot of superheat at the onset of nucleate boiling both in the upflow and in the downflow.     

Experiments on microgravity boiling heat transfer by using transparent heaters

To clarify the relation between the liquid–vapor behavior and the heat transfer characteristics in the boiling phenomena, the structures of transparent heaters were developed for both flow boiling and pool boiling experiments and were applied to the microgravity environment realized by the parabolic flight of aircraft. In the flow boiling experiment, a transparent heated tube makes the heating, the observation of liquid–vapor behavior and the measurement of heat transfer data simultaneously possible. The heat transfer coefficient in the annular flow regime at moderate quality has distinct dependence on gravity provided that the mass velocity is not so high, while no noticeable gravity effect is seen at high quality and in the bubbly flow regime. The measured gravity effect was directly related to the behavior of annular liquid film observed through the transparent tube wall. In the pool boiling experiment, a structure of transparent heating surface realizes both the observation of the macrolayer or microlayer behavior from underneath and the measurements of local surface temperatures and the layer thickness. It was clarified in the microgravity experiments that no vapor stem exists but tiny bubbles are observed in the macrolayer underneath a large coalesced bubble at high heat flux. The heat flux evaluated by the heat conduction across the layer assumes less than 30% of the total to be transferred. The evaporation of the microlayers underneath primary bubbles just after the generation dominates the heat transfer in the microgravity, not only in the isolated bubble region but also in the coalesced bubble region.     

竖直管道内间歇式两相流动沸腾特性分析

自然循环或重力注水过程的热功率、冷却剂流量等操作条件较小,易出现各种流动不稳定现象,影响核反应堆事故的发展进程,间歇式流动沸腾现象就属于其中的一种。以去离子水为工质,采用2×2加热棒束,对内径为32 mm竖直通道内的间歇式流动沸腾现象进行了实验研究,分析了不同热流密度下间歇式流动沸腾不稳定现象的变化规律,讨论了热流密度对间歇式沸腾周期的影响。结果表明,在一定的热流密度条件下,当加热通道内流体达到饱和并过热时,会发生周期性地剧烈喷涌及冷液回流现象,期间伴随泡状流、弹状流、搅混流及环状流等多种流动形态;间歇喷涌周期取决于沸腾停滞时间,随热流密度的不断增大,沸腾停滞时间缩短,间歇喷涌周期也缩短。当热流密度增大到一定程度时,间歇式流动沸腾现象消失,从而转变为另一种两相流动不稳定现象。     

The effect of flow direction and magnitude on CHF for low pressure water in thin rectangular channels

Critical heat flux (CHF) at low flow condition can become important in an MTR-type research reactor under a number of accident conditions. Regardless of the initial stages of these accidents, a condition which is basically the decay heat removal by natural convention boiling can develop. Under such conditions, burnout may occur even at a very low heat flux. In view of this, the CHF at low-flow-rate and low-pressure conditions has been studied for water flowing in thin rectangular channels.Experiments were carried out with two types of rectangular test sections, namely, the one heated from one wide side and the other heated from two opposite sides. In order to observe the effects of gravity, CHF was measured both in upflow and downflow. The CHF at complete bottom blockage was also studied.The results indicate that burnout can occur at a much lower heat flux than pool-boiling CHF or than predicted by the conventional correlations. There was observed a minimum CHF at complete bottom blockage and at very low downflow. The low CHF at very low downflow appears to be due to the stagnation of the bubble in the heated section. This fact indicates that special care should be taken in analyzing the boiling phenomenon which occurs when the coolant flow is very low in a low pressure system.     

Experimental Study of Differences in DNB Heat Flux between Upflow and Downflow in Vertical Rectangular Channel

Experiments were carried out on DNB (Departure from Nucleate Boiling) heat flux for both upflow and downflow in a rectangular vertical channel simulating a subchannel in the fuel element of the research reactor JRR-3, which is scheduled to be upgraded at 20MWt with 20% low enriched uranium (LEU) fuel. The experiments were carried out for the conditions of pressure and velocity which are important for the safety design of the JRR-3. With the investigation of the data of the present experiment along with already existing data for both rectangular channels and other channels, a scheme of DNB heat flux correlations was obtained for upflow and downflow. With the all available data, the errors of the correlations adopted in this scheme were so evaluated as to utilize these results in the safety analysis of the JRR-3. This scheme includes a new correlation of DNB heat flux for upflow and the identification of a region of high coolant flow rate where no remarkable differences in the DNB heat flux are observed between upflow and downflow. This scheme is considered to be applicable not only to the rectangular channels of the JRR-3 but also to other channels.     

Natural Convection Sodium Boiling Experiments in 37-Pin Bundle Geometry

Decay heat removal capability under boiling condition was studied using an LMFBR fuel subassembly mockup loop. The sodium flow was driven by natural convection through the loop in which was installed a 37-pin bundle heated electrically over a length of 45 cm.

The heat flux furnished by the pins was increased stepwise, upon which the two-phase flow regime changed from bubble to slug flow and then to annular or annular mist flow. Dryout occurred even in slug flow regime, but only momentarily, and permanent dryout was not observed before establichment of annular flow. A suitable criterion for permanent dryout is considered to be 0.5 average exit sodium vapor quality. The results indicated that upon occurrence of sodium boiling, the coolability of fuel subassembly would be maintained by natural convection after reactor shutdown.     

液钠沸腾两相流流型与临界热流密度(CHF)机理实验研究

通过大量的液态金属钠临界热流密度 (CHF)的实验研究 ,结合液钠两相传热流动特性及液钠的物性特点 ,分析了起始沸腾流型 ,泡状流 ,块状流 ,环状流和双向环状流的热工水力特性 ;并从实验结果出发 ,深入分析了液钠发生临界热流密度时的气泡爆炸和液膜撕裂或局部蒸干的两种传热恶化机理     

Experimental investigation of heat transfer for supercritical pressure water flowing in vertical annular channels

An experiment has recently been completed at Xi’an Jiaotong University (XJTU) to obtain wall-temperature measurements at supercritical pressures with upward flow of water inside vertical annuli. Two annular test sections were constructed with annular gaps of 4 and 6 mm, respectively, and an internal heater of 8 mm outer diameter. Experimental-parameter ranges covered pressures of 23-28 MPa, mass fluxes of 350-1000 kg/m²/s, heat fluxes of 200-1000 kW/m², and bulk inlet temperatures up to 400 °C. Depending on the flow conditions and heat fluxes, two distinctive heat transfer regimes, referring to as the normal heat transfer and deteriorated heat transfer, have been observed. At similar flow conditions, the heat transfer coefficients for the 6 mm gap annular channel are larger than those for the 4 mm gap annular channel. A strong effect of spiral spacer on heat transfer has been observed with a drastic reduction in wall temperature at locations downstream of the device in the annuli. Two tube-data-based correlations have been assessed against the experimental heat transfer results. The Jackson correlation agrees with the experimental trends and overpredicts slightly the heat transfer coefficients. The Dittus-Boelter correlation is applicable only for the normal heat transfer region but not for the deteriorated heat transfer region.