Development of Steam Separator Performance Analysis Code and Its Validation, (II) Carryover Experiments

Abstract

The purpose of the study is to collect data in order to make models that are applicable to calculate carryover droplets that are generated in and flow out of a steam separator. Various effective tests relevant to separation mechanisms in the separator have been conducted with a full-scale steam separator under atmospheric pressure. Separation behaviors for the top of the riser of the separator and for corrugated-separator were clarified and correlated by the experiment. Distinct patterns about the separation at the corrugated-separator, the separation of discharged droplets by gravity, and the separation of droplets by a screen dryer that is used to dry up the steam were also measured with the facility using a full-scale separator in a vessel simulating the flow area of ATR under the high-pressure and high-temperature condition for various water levels. Each separation data were correlated under the condition of maximum steam and liquid flow rates of 7 and 30.5 kg/s, respectively. Liquid droplets containing a small amount of LiOH at several positions were sampled together with steam by iso-kinetic probes, and the amount of carryover was analyzed in PPT range by the chemical analysis of condensed steam. As a result, basic data for separation mechanisms were obtained, and maximum capacity of the separator was estimated.     

Characteristics of two-phase condensing flow by visualization using computed image processing

The mechanics of the condensing behavior of vapor bubbles in a subcooled bulk flow is complicated and influenced by both heat and mass transfer. To examine the characteristics of such thermal-nonequilibrium two-phase flow, experimental and analytical researches have been made. In the experiment, the movement of each vapor bubble in a flowing channel was recorded on video tapes and analyzed by an image processing system. As result, the distributions of void fraction along the test section were obtained. In the analysis, a simple analytical model was introduced to predict the distributions of void fraction and liquid subcooling temperature. By considering the rate of vapor condensation along the flow direction, the differential equation of energy balance between two phases was obtained. Integration of this equation yielded the void fraction and bulk liquid subcooling at any position. The condensation rate was estimated as a function of the local liquid subcooling, interfacial area and mass velocity. Finally, a close fit between calculated results and experimental data was obtained.     

Network Analysis of Turbine and Feedwater Systems of the ‘Fugen’ Nuclear Power Plant

The present study describes the thermal-hydraulic network analysis of the turbine and feedwater systems of the ‘Fugen’ reactor. Turbines, feedwater heaters, and corresponding piping systems are modeled using the network calculation code NETFLOW++ and thermal-hydraulic conditions are calculated using the coupled numerical model. As a result of the calculation, distributions of important characteristics of the single-phase flow and two-phase flow in the piping such as pressure and void fraction are clarified. Flow patterns in the piping were investigated using the calculated result. It was found that the state of the coolant in the drainpipe changes from saturated liquid at the inlet to a two-phase flow with a large void fraction at the connection to the feedwater heaters. This is attributed to the pressure difference between the inlet and outlet of the drainpipes. Even the drainpipe from the moisture separator to the shell of feedwater heater #4 shows a similar behavior, and the flow pattern changes from single phase to slug flow. The steam quality in the extraction line is very high, although a large number of droplets are contained in the flow. Contrary to expectation, these droplets do not completely evaporate in spite of the low-pressure conditions.     

Development of a Numerical Analysis Method for Cross-sectional Void Fraction Distribution in Subchannels of Nuclear Fuel Assemblies

Detailed information about the void fraction distribution in fuel assemblies is increasingly important with the development of high burn-up fuels. A numerical method has been developed for the steady cross-sectional void fraction profile in fuel assemblies using a marching method in the axial direction, considering cross-flows due to lift forces, void diffusion and momentum balance. Uniform pressure in a cross section was assumed under the dominant vertical flow and the secondary lateral flow condition in each subchannel. The merit of this simplified method is its high-performance computation using many BFC meshes for expression of complex void fraction and velocity distributions inside the subchannels. The calculated results were compared with the observed void distributions obtained with X-ray computed tomography in the NUPEC tests of full-scale advanced BWR fuels. The comparison showed the capability of this method for predictions of overall void fraction distributions inside the subchannels. This method will provide a good tool for void fraction profile prediction in high burn-up fuels, while future studies for reliable correlations of lift forces are required over a wide range of flow conditions.     

Void distribution and bubble motion in bubbly flows in a 4×4 rod bundle. Part I: Experiments

Lack of local void fraction data in a rod bundle makes it difficult to validate a numerical method for predicting gas–liquid two-phase flow in the bundle. Distributions of local void fraction and bubble velocity in each subchannel in a 4×4 rod bundle were, therefore, measured using a double-sensor conductivity probe. Liquid velocity in the subchannel was also measured using laser Doppler velocimetry (LDV) to obtain relative velocity between bubbles and the liquid phase. The size and pitch of rods were 10 and 12.5 mm, respectively. Air and water at atmospheric pressure and room temperature were used for the gas and liquid phases, respectively. The volume fluxes of gas and liquid phases ranged from 0.06 to 0.15 m/s and from 0.9 to 1.5 m/s, respectively. Experimental results showed that the distributions of void fraction in inner and side subchannels depend not only on lift force acting on bubbles but also on geometrical constraints on bubble dynamics, i.e. the effects of rod walls on bubble shape and rise velocity. The relative velocity between bubbles and the liquid phase in the subchannel forms a non-uniform distribution over the cross-section, and the relative velocity becomes smaller as bubbles approach the wall due to the wall effects.     

棒束定位格架两相CFD模拟方法研究

考虑气泡合并分裂,采用MUSIG模型,对3×3格架内空气-水两相分布进行计算流体力学(CFD)数值模拟研究发现,计算对入口两相分布预计不敏感,但对气泡直径大小敏感;在定位格架下游不远处,空泡份额分布由较小直径气泡起主导作用,格架下游较远处,空泡份额分布由较大直径气泡起主导作用。考虑空气-水两相流量、几何条件和压力对气泡直径的影响,本文提出针对棒束定位格架的数值模拟气泡最大直径设置关系式,并对模型选取和模拟方法给出建议。计算表明空泡份额分布曲线形状与峰值均和实验符合较好,该模拟方法能合理预测复杂通道两相数值分布。     

Void Transport after Steam Generator Tube Leakage in Natural Circulation Lead-cooled Fast Reactor

For lead-cooled fast reactors, steam generator tube leakage and/or rupture (SGTL/R) is one of the safety issues. During SGTL/R, high-pressure water from secondary side of main heat exchangers is injected to lower-pressure primary side. It may have significant negative impact on the integrity of structures nearby, the flow and heat transfer capabilities of the primary system and the reactivity of the core. The transport of steam bubbles and subsequent void accumulation in the primary system of SNCLFR-100 were addressed in the paper. Based on ANSYS FLUENT, Lagrangian tracking of steam bubbles (voids) in Eulerian flow field was performed to identify the locations and traces of steam bubbles after SGTL. The core safety characteristics under SGTL/R accident were also evaluated. The results show that the leakage location, bubble size and the liquid contamination all have influence on the void transport. For the contaminated primary system of a lead-cooled fast reactor, the occurrence of SGTL with a low leakage location may significantly affect the normal operation of the reactor.     

自然循环铅冷快堆蒸汽发生器泄漏事故下的气泡迁移

蒸汽发生器传热管泄漏/破裂事故是核电厂平稳运行的安全问题之一,对铅冷快堆而言,该事故发生后,二回路的高压水迅速进入一回路,会对蒸汽发生器传热管邻近的结构、一回路的流动、一回路换热乃至堆芯的反应性产生较大影响。本文针对SNCLFR-100小型自然循环铅冷快堆,对破裂后气泡的迁移以及在反应堆的积聚进行研究,基于ANSYS FLUENT,利用欧拉-拉格朗日方法对泄漏后气泡的位置和轨迹进行了追踪,并对事故下的堆芯安全进行了一定的评估。研究表明,破裂位置、气泡尺寸以及冷却剂纯净度均会对一回路气泡的迁移产生较大的影响,当一回路液态铅含有较多杂质时,蒸汽发生器较低位置发生的泄漏事故会产生相当大的系统气泡积聚和堆芯气泡累积,从而对反应堆的正常运行产生显著影响。     

旋叶分离器分区域疏水性能研究

旋叶分离器是蒸汽发生器中气水分离器组件的关键部件之一,保障蒸汽发生器为汽轮机提供高干度的蒸汽。气液混合物进入分离器后,液体逐渐分离,并通过入口回流、下降通道、切向口等区域疏水。为研究各区域疏水性能,建立了可视化旋叶分离器冷态试验回路,采用常温常压空气 水作为工质,针对直径140 mm、旋叶倾角30°的分离器开展分离性能试验。测试获得了空气表观速度在5~13 m/s区间内入口回流、下降通道、切向口对分离水比例变化的影响规律。研究发现,低气相表观速度时入口回流和下降通道承担主要的分离水比例,中等气相表观速度（7~9 m/s）时下降通道承担主要分离水比例,高气相表观速度时切向口起重要的分离水作用。     

流体物性对冷中子源冷包截面含气率的影响模化研究

采用差压法测出了冷中子源模拟冷包中环形通道内不同层面的平均含气率,发现液相的表面张力与密度的比值是决定截面含气率的主要物性因素;此比值越大,含气率越低。氟利昂113的表面张力与密度的比值比液氢的相应值低。故可以选用氟利昂113作为模拟工质对中国先进研究堆冷中子源中气液两相氢循环的含气率进行模拟研究,且试验结果将偏于保守。     

Study on intermittent flow behavior in a vertical channel under low-pressure condition

The intermittent flow behavior in a vertical annulus under a low-pressure condition was experimentally studied using a scaling experiment facility. The temperature and pressure variations in the channel had been obtained under the heat load ranging from 0 to 2.0 kW, initial subcooled water temperature ranging from 50 to 90 °C and length–diameter ratio ranging from 1.6 to 50. The effects of the heat load and length–diameter ratio of channel on the flow characteristics were investigated in detail. The experimental results showed that the steam bubbles erupted more frequently and regularly at a high heat load. The intermittent flow period decreased with increase of the heat load and aspect ratio. Based on the mechanism analysis, an empirical model considering the steam oscillation and the vapor–liquid interface rupture based on the experimental data was proposed. It was found that the accumulated steam basically increased linearly. The oscillation of the pressure and velocity decreased gradually with continuous steam accumulation. The Reynolds number of the liquid within the rising section was very small at the stagnation state since there was no forced circulation flow. Finally, a blockage was engendered in the pipeline with the steam accumulated.     

Experimental Study on Gas Carry-Under in Liquid Down Flow from a Two-Phase Mixture

The gas carry-under characteristics in liquid down flow from a two-phase mixture flow have been studied for various flow parameters, based on experiments with a small scale air- water system simulating the concept of a natural circulation BWR with no separators. For high void fraction in the riser, as the liquid superficial velocity jf increased to 0.17 m/s, the void fraction in the lower part of the downcomer αd increased sharply due to the descent of comparatively large bubbles (diameter: about 4–6mm). In the region of jf> 0.17m/s, on increasing jf, the void fraction αd increased until it reached a maximum value at jf.3. For liquid descending velocities higher than 0.3 m/s, αd became almost constant and the level of the mixture above the riser had little effect on the void fraction ad due to the phase separation of the large bubbles formed by bubble coalescence in the upper part of the downcomer. The void fraction αd increased as the void fraction αr increased until bubble coalescence occurred in the upper part of the downcomer, and αd became constant and independent of αr after the occurrence of bubble coalescence. Under the conditions of high void fractions in the riser, 0.4<αr<0.64 (upper limit of the tests), and high liquid descending velocities in the down-comer, 0.3 m/s<jf.<0.4 m/s (upper limit of the tests), the void fraction αd was represented by a dimensionless number (G = η⁴ g/σ³ pf) and by the upper limit of void fractions in bubbly flow, αd=0.3.     

Mockup tests for developing the CARR-CNS with a two-phase thermo-siphon loop

The cold neutron source (CNS) is a facility to increase cold neutrons by scattering thermal neutrons in liquid hydrogen or deuterium around 20 K. For extracting a stable cold neutron flux from the CNS, the liquid quantity in the moderator cell should be maintained stably against disturbance of nuclear heating. The China Institute of Atomic Energy (CIAE) is now constructing the China Advanced Research Reactor (CARR: 60 MW). and designing the CARR-CNS with a two-phase thermo-siphon loop consisting of a condenser, two moderator transfer tubes and an annular cylindrical moderator cell. The mock-up tests were carried out using a full-scale loop with Freon-113,for validating the self-regulating characteristics of the loop, the void fraction less than 20% in the liquid hydrogen of the moderator cell, and the requirements for establishing the condition under which the inner shell has only vapor. The density ratio of liquid to vapor and the volumetric evaporation rate due to heat load are kept the same as those innormal operation of the CARR-CNS. The results show that the loop has the self-regulating characteristics and the inner shell contains only vapor, while the outer shell liquid. The local void fraction in the liquid increases with increasing of the loop pressure.     

Mitigation Technologies for Damage Induced by Pressure Waves in High-Power Mercury Spallation Neutron Sources (III)—Consideration of the Effect of Microbubbles on Pressure Wave Propagation through a Water Test—

The effects of microbubbles dispersed in a liquid on a high-rising-rate pressure wave were experimentally investigated with water. Intense, high-rising-rate pressure waves with a rise time of about 1.5 ms were produced by a spark discharge in water, and gas microbubbles were produced by two different bubble generators. Particular attention was focused on the attenuation effect of microbubbles on propagating pressure waves. The dependence of the attenuation effect on the radius and void fraction of the microbubbles was carefully examined. It was found that when the microbubbles are sufficiently small (e.g., about 50 μm in peak radius), the amplitude of wall vibration induced by the spark-induced pressure wave is dramatically decreased with an increase in void fraction. The present study provides strong experimental evidence that microbubbles can act as a strong absorber for high-rising-rate pressure waves as recently predicted numerically.     

竖直窄矩形通道内空气-水两相流中气弹运动速度研究

以空气和水为工质,应用高速摄像仪,对竖直窄矩形通道(3.25 mm×40 mm)内气液两相弹状流进行了可视化实验研究。气、液相表观速度分别为0.1~2.51 m/s和0.16~2.62 m/s,工作压力为常压。实验中发现窄矩形通道内弹状流与圆管中存在较大差别,气弹多发生变形,高液相流速时变形更为严重。窄边液膜含气量较高,在高液相流速时窄边液膜不下落,宽边液膜中含有由气弹头部进入和气弹尾部进入的气泡。气弹速度受气弹头部形状和宽度影响较大,受气弹长度影响较小。气弹速度可由Ishii & Jones-Zuber模型计算,但在低液相折算速度时偏差较大,其主要原因为漂移速度计算值较实验值偏小。     

Evaluation of Delayed Neutron Fraction β for Thermal Fission of U-235 Based on Integral Experiments Using SHE

Countercurrent gas-liquid flow is theoretically and experimentally evaluated for a boiling system simulating a BWR core. In a single channel, flow patterns are determined from the mass balance equations and pressure drop under steady state conditions is calculated for each flow pattern using a drift flux model, where the distribution parameter and drift velocity are correlated as functions of void fraction and hydraulic diameter from void fraction data. The calculated pressure drop shows a similar trend to that of the data for the effects of bypass leak flow rate and heater power. Countercurrent behavior in three boiling channels under slow transient conditions is also predicted from the single channel characteristics and close agreement is obtained between the predicted and experimental results. The results show that steam up-flow or cocurrent up-flow easily occurs in a channel with low pressure drop, namely, with a large entry orifice, high power or low bypass leak flow rate.     

Influences of physical properties of two-phase mixture on voidfraction in an annular vessel

To keep the void fraction of two-phase hydrogen in the moderator cell of the cold neutron source (CNS) of China Advanced Research Reactor (CARR) to a specified range, an annular vessel with the same size as the actual moderator cell was used as test section. Deionized water and alcohol, sucrose, and sodium chloride solutions with different concentrations were used as working fluid to find out influences of physical properties, such as density, viscosity and surface tension, of the two-phase mixture on void fraction. The tests proved that the ratio of surface tension to density of liquid phase has great influence on void fraction: the larger the ratio, the smaller the void fraction. Since the ratio of surface tension to density of Freon 113 is lower than that of liquid hydrogen, Freon 113 can be used as a working fluid to study the void fraction in the two-phase hydrogen thermosiphon loop in the CNS of CARR and the results will be conservative.     

A unified model considering force balances for departing vapour bubbles and population balance in subcooled boiling flow

Population balance equations combined with a three-dimensional two-fluid model are employed to predict subcooled boiling flow at low pressure in a vertical annular channel. The MUltiple-SIze-Group (MUSIG) model implemented in CFX4.4 is extended to account for the wall nucleation and condensation in the subcooled boiling regime. A model considering the forces acting on departing bubbles at the heated surface is formulated. This model provides the capacity of complex analyses on the bubble growth and departure for a wide range of wall heat fluxes and flow conditions.Comparison of model predictions against local measurements is made for the void fraction, bubble Sauter mean diameter and gas and liquid velocities covering a range of different mass and heat fluxes and inlet subcoolings. Good agreement is achieved with the local radial void fraction, bubble Sauter mean diameter and liquid velocity profiles against measurements. However, significant weakness of the model is evidenced in the prediction of the vapour velocity. Work is in progress to circumvent the deficiency of the MUSIG boiling model by the consideration of additional momentum equations to better represent the momentum forces acting on the range of bubble sizes in the bulk subcooled liquid.     

Numerical approach on the performance prediction of a gas–liquid separator for TMSR

In this paper, the numerical simulation for the gas–liquid flow in a separator applied in the fission gas removal system for thorium molten salt reactor was investigated. The numerical model was established in the frame of Eulerian–Eulerian approach, in which the modeling of the forces acting on the bubbles was introduced. Based on the model, numerical simulations with three flow rates were carried out. Three key parameters (the pressure loss, the separation length, the liquid entrainment ratio) concerned with the separation performance were compared between the numerical results and the experimental data, the results indicate that the calculated results agree well with the experimental data. Hence, the numerical approach shows a promising tool for the performance prediction and the optimization of the gas–liquid separator.     

自然循环两相漂移流理论分析模型的建立

以5MW核供热堆试验回路（HRTL－5）为物理原型，从理论上推导了加热段欠热沸腾，上升段入口附近的冷凝，上升段闪蒸以及气空间压力平衡等关系式，并考虑了热力不平衡等因素的影响，建立了一个完整的四方程漂移流模型，该模型适用于分析低压低干度自然循环系统的流动问题，通过引入凝边界层，首次解决了两相流动过程中汽泡在过冷水中的冷凝问题，对汽空间则推导了压力及液位的约束方程，方程组可采用积分方法求解，并利用Rung-Kutta-Verner方法求解所得的微分方程组。