液态金属内单个气泡上升行为的MPS法数值模拟

液态金属冷却核反应堆采用气泡泵的概念设计来提升堆芯自然循环能力。液态金属内气液两相流动特征将直接影响核反应系统一回路的自然循环能力及堆芯安全。本研究通过采用移动粒子半隐式（MPS）方法,对液态金属中单个上升气泡的气泡动力学行为进行数值模拟。分析了铅铋合金中3种初始直径不同的单个氮气泡在上升过程中的气泡形状和速度的变化趋势;对比了初始直径相同的单个氮气泡在液钾、液钠、铅铋合金、钾钠合金和锂铅合金5种液态金属中的上升行为;同时将模拟得到的气泡形状与Grace经验关系图进行了对比,验证了MPS方法数值模拟结果的正确性。     

铅铋合金冷却反应堆内气泡提升泵提升自然循环能力的理论研究

液态金属冷却核反应堆采用气泡提升泵的概念设计来提升堆芯自然循环能力。液态金属和惰性气体两相流动特性显著影响自然循环能力和系统安全性。本工作对铅铋合金冷却反应堆中气泡提升泵提升自然循环能力进行数值模拟研究。基于漂移流模型,采用空泡份额预测模型和摩擦压降预测模型分析了气体质量流量、质量含气率、气泡直径、上升管道高度对气泡提升泵提升自然循环能力的影响。结果表明：泡状流区域中,对于给定的气体质量流量,随着充入气泡直径减小,自然循环能力呈增加趋势。在泡状流、弹状流、乳状流和环状流等流型中,随着气体质量流量、质量含气率增加,自然循环能力先增大后减小。随着上升管道高度增加,自然循环流量增大。可见,流动参数显著影响堆芯热工水力特性。现有工作有助于优化带有气泡提升泵的自然循环冷却系统设计。     

静水中单气泡运动特性实验研究

气泡运动特性是两相流研究的重要领域之一。本文基于图像识别技术对静水中单气泡的上升过程及其动力学特性进行研究，分析了不同等效体积直径下气泡上升过程中轨迹、变形程度、瞬时速度等参数。研究结果表明：椭球型单气泡上升过程中横向振荡幅度峰值与横向速度峰值呈正比，瞬时纵横比与瞬时纵向速度呈反比；单气泡横向运动位置、纵横比、上升速度具有周期性变化的现象。通过定量分析气泡末速度与等效体积直径的关系，对已有预测关系式进行了比较和评价，并基于实验数据拟合得到了一个更为精确的显式末速度关系式。     

Experimental Investigation on Characteristic of Single Bubble Motion in Stagnant Water

The bubble motion characteristic is one of the most important fields in two-phase flow research. Based on the image recognition method, the characteristic and dynamics of single bubble rising process in stagnant water were analyzed. The trajectory, deformation extent and instantaneous terminal velocity of single bubble rising process with different equivalent volume diameters were analyzed. The research results show that the peak value of lateral oscillation amplitude is proportional to the peak value of lateral velocity. The instantaneous aspect ratio is inversely proportional to instantaneous axial velocity. The phenomenon that the lateral motion position, aspect ratio and rising speed of a single bubble has periodical change. Based on the quantitative analysis of the relationship between bubble terminal velocity and bubble equivalent volume diameter, the existing prediction formulas were compared and evaluated. A more accurate explicit terminal velocity correlation based on experimental data was obtained.     

轴流铅铋泵流场分析及优化

铅铋泵作为铅铋冷却快堆一回路的关键输送设备,其安全运行对铅铋冷却快堆的安全至关重要。液态铅铋合金在泵内流动特性对泵的长期安全运行有重要影响,为了研究轴流铅铋泵泵内流场,通过Workbench/BladeGen软件建立了主泵叶轮模型,在ANSYS CFX软件中数值模拟泵内流场,并根据数值模拟结果改进了导叶片厚度,优化了动叶片翼型出口角,从而改善泵内流场。研究结果表明,铅铋泵叶片型线出口附近角度变化过快会导致叶片压力分布不均匀,产生局部高压的现象,进而可能造成更严重的冲蚀。优化导叶片厚度以及动叶片出口液流角后,泵内流场整体迹线较为平稳,导叶片出口处铅铋合金流速可以维持在1.8 m/s左右。      

Study on flow characteristics in gas-molten metal mixture pool

As part of basic research on the flow characteristics of a two-phase mixture pool under severe accident of fast breeder reactor (FBR), visualization and measurement of nitrogen gas-molten lead/bismuth two-phase flow in a rectangular pool were performed by using the neutron radiography technique. Measurements of drag coefficient of a single bubble and bubble shape regime showed that the relationship between the shape, size and the rising velocity of a single isolated nitrogen bubble in the molten lead/bismuth was not much different from that for an ordinary one. Appropriate correlation for drift velocity and drag coefficient between phases were recommended based on the drift flux correlation of measured pool void fraction. One- and two-dimensional analyses were performed by using a next generation computational code for safety analysis of severe accident of FBRs, SIMMER-III with various drag coefficient models. It was revealed that Kataoka–Ishii’s equation was suitable basically for estimation of drift velocity, namely, drag force between phases.     

铅铋自然循环热工水力分析程序开发

铅铋堆内冷却剂的自然循环对于反应堆的正常运行以及事故工况下的堆芯热量导出均至关重要,相关热工水力分析工作对于支持设计及安审均有重要意义。通过对铅铋堆内一回路系统内主要部件,包括堆芯、热交换器、管道等建立热工水力物理模型,开发了适用于铅铋自然循环瞬态过程模拟的热工水力分析程序,并利用铅铋自然循环回路内开展的自然循环启动实验、功率台阶影响实验等的结果进行了程序的初步验证。结果表明,程序计算得到的结果与实验结果符合较好,能够较好模拟铅铋自然循环的瞬态过程。该程序可以为铅铋堆研发过程中自然循环热工水力分析工作提供支持。     

Effects of Mass Transfer Resistances in Fluid and Permeable Walls on Hydrogen Permeation in Tube-Type Metal Window

The gas lift pump concept based on the bubbling of an inert gas into the primary reactor coolant to enhance natural circulation is currently considered in a number of PbBi-cooled reactor concepts. Thus, the analysis of available void fraction data and the development of two-phase heavy liquid metal/gas flow calculational models have become an important issue in the study of advanced nuclear reactor systems. In the absence of the detailed two-phase flow information needed to develop a flow regime map and the associated interfacial relations, drift-flux models have often been used in the thermal-hydraulic analysis of nuclear and other systems. Accordingly, we consider, in the current paper, the analysis of five sets of experimental data with different geometries, working fluids, flow rates and void fraction ranges, with a view to obtaining a best fit to the data in the form of a drift-flux model. The results of the analysis show that, for systems with flowing fluid, it is possible to represent the heavy liquid metal void fraction data in the form of a drift-flux correlation with a residual error of as low as 0.016, thus offering an improvement over existing void correlations.     

Liquid metal corrosion/erosion investigations of structure materials in lead cooled systems: Part 1

In future lead cooled nuclear power systems the heavy liquid-metal pump will be placed in the hot temperature region of the reactor. This combines corrosion problematic at elevated temperatures with erosion at the impeller of the pump. Several steels designed for conventional mechanical high loaded pumps and the SiSiC have been tested in oxygen containing stagnant lead (10⁻⁶ and 10⁻⁸ wt%) at 550 and 600 °C for 2000 and 4000 h in the COSTA-facilities. Only SiSiC shows no influence by the liquid metal. No dissolution attack occurs at cast iron steels but inner oxidation takes place. NORILOY shows no dissolution attack. All other steels are attacked by liquid lead at one of the conditions. To evaluate the erosion-corrosion attack a new test facility allowing velocities of the lead of up to 20 m/s was designed and constructed. With a CFD-code the behaviour and flow velocity of the lead are simulated.     

Subchannel analysis of fuel assemblies of a lead–alloy cooled fast reactor

Lead–alloy cooled fast reactor is one of the six Gen-IV reactors. It has many attractive features such as excellent natural circulation performance, better shielding against gamma rays or energetic neutrons and potentially reduced capital costs. A natural circulation lead–alloy cooled fast reactor with 10 MWth is under design in China (hereafter called LFR-10MW). Fuel assemblies thermal hydraulic analysis is of vital importance for a successful design. A subchannel analysis code with flow distribution model was used to carry out the thermal hydraulic analysis. This work briefly gave the thermal-hydraulic design for the LFR-10MW and analyzed the thermal-hydraulic characteristics under steady-state condition using the subchannel analysis code. Whole core analysis was performed to locate the hottest fuel assembly using the code. The hottest fuel assembly was analyzed to obtain the cladding temperature, fuel temperature and coolant velocity. The maximum cladding temperature, the maximum fuel center temperature and the maximum coolant velocity are all below the design constraints. These results imply that the thermal-hydraulic design of LFR-10MW is feasible.     

基于扩散界面法的液态铅铋合金中气泡上升行为模拟

基于扩散界面法,对单个氮气气泡在液态铅铋合金内从静止到充分发展整个过程中的动力学行为进行数值模拟,得到气泡形变特性和气泡上升速度随时间的变化关系,将模拟结果与Grace经验关系图对比,发现模拟得到的气泡形变结果在Grace经验关系图中均可找到且很好地吻合,从而验证了扩散界面法在模拟液态铅铋合金中气泡上升行为的可行性和准确性。同时基于界面扩散法的模拟,对比了5种不同初始直径的氮气泡在液态铅铋合金中的上升行为,发现初始直径较小的气泡在上升过程中扰动会更剧烈,初始直径较大的气泡在上升过程中易发生分裂现象。     

Aspects of minimizing steel corrosion in liquid lead-alloys by addition of oxygen

Minimizing steel corrosion in liquid lead-alloys by addition of oxygen requires devices for efficient oxygen transfer and reliable oxygen sensors. The accuracy of electrochemical oxygen sensors is analyzed using theoretical considerations and results from experiments in stagnant lead–bismuth eutectic (LBE). Additionally, the feasibility of gas/liquid oxygen-transfer and the long-term performance of electrochemical sensors in flowing liquid metal are addressed on the basis of the operating experience of the CORRIDA loop, a facility for testing steels in flowing LBE.     

铅铋自然循环回路热损效应分析

在液态金属自然循环回路的计算分析过程中,已有研究一般忽略散热损失,常导致计算结果与实验结果有较大的区别。为研究散热损失对液态金属自然循环回路稳态特性的影响,利用MATLAB/Simulink编制了含有散热损失模型的铅铋自然循环回路计算程序,并用实验结果进行了验证。利用该程序,分析了不同热功率、中间热交换器二次侧流量和环境温度下散热损失对自然循环回路稳态参数的影响。计算结果表明:通过减小散热损失可提高回路的自然循环流量;当二次侧流量较小时,散热损失对循环流量的影响更为明显;通过增加二次侧流量或适当增加热功率可减小散热量占总热功率的比例,提高热量利用率;当二次侧流量不变时,不同热功率下环境温度对回路的自然循环流量的影响不明显,但热量利用率会随环境温度的升高而增加。     

Models of liquid metal corrosion

In the present study, models for liquid metal corrosion are reviewed and their applications in nuclear reactor engineering are discussed. The paper presents mathematical analysis of liquid metal corrosion, including species transport in solid steels, in flowing liquid metals, and mass exchange at liquid/solid interface. The survey illustrates the mechanisms of the liquid metal corrosion and sets up a system to calculate the corrosion rate and to study the corrosion species distributions in the solid and liquid metal/alloys. Both light liquid metal/alloy (sodium and sodium-potassium) and heavy liquid metal/alloy (liquid lead and lead-bismuth) are considered. Oxygen effects on liquid metal corrosion are also discussed. For liquid sodium and sodium-potassium the corrosion rate increases with increasing oxygen concentration, while for liquid lead and lead-bismuth it is reasonable to produce a protective oxide layer using an oxygen control technique which can mitigate the corrosion rate significantly. Finally, the corrosion-oxidation interaction in liquid lead and lead-bismuth are discussed.     

双铅铋回路自然循环瞬态响应分析

为了研究液态金属双自然循环回路功率变化时的瞬态响应和回路的稳定性,利用SIMULINK在单个铅铋自然循环回路的基础上建立了双自然循环回路模型,并用FLUENT程序进行计算验证,分析了不同升功率和降功率过程中回路的流速振荡及双回路的稳定性。计算结果表明:延长升功率时间可减小流速振荡,且当升功率速率先慢后快时更有利于减小流速振荡和缩短流动到达稳定所需的时间;对于双自然循环回路,当降功率速率不同时,回路的流速均未发生明显的振荡,降功率速率对回路的稳定性影响并不显著。双自然循环回路比单回路更不稳定,通过减小中间换热面的热阻能有效提高双回路的稳定性。     

Numerical investigation of bubble wake properties in the moving liquid with LES model

Wake flow caused by the relative motions between bubble and liquid phase influences bubble motions and enhances turbulent properties in the liquid phase. This phenomenon has been stressed for a better understanding of bubbly flow. In this paper, large eddy simulation (LES) is performed to simulate a single bubble rising in the moving liquid, with volume of fluid (VOF) method to capture the interface movements between bubble and liquid phase. The simulation results are firstly compared with the numerical and experimental data from the literature. A good agreement demonstrated the capability of the employed computational fluid dynamics (CFD) approach to predict turbulent properties in the liquid phase and capture interface movement as well as its induced wake flow. Consequently, the dynamic behaviors of a single bubble rising in the moving liquid were investigated. An ensemble averaged has been employed to evaluate the velocity distribution composed by wake velocity and liquid velocity quantitatively as well as the velocity fluctuations enhanced by bubble motion. Their dependency was also evaluated based on a systematic CFD simulation which covers a wide range of liquid velocity. With comparisons of the single phase flow, the influence from the existence of bubble on turbulent properties was determined.     

Testing and qualification of CIRCE instrumentation based on bubble tubes

The paper is focused on the development and the qualification of the instrumentation suitable for the level and the differential pressure measurements in the experimental facility CIRCE. This large-scale facility is installed at the ENEA Brasimone Centre for studying the fluid-dynamic and operating behavior of ADS reactor plants. This a rather challenging objective since the facility adopts the molten lead bismuth eutectic (LBE) alloy as a coolant and, at present, instrumentation qualified for operating under such conditions is scarce or does not exist. Bubble tubes have been installed in CIRCE to transfer pressure signals from the LBE to differential pressure cells operating with gas at room temperature. The bubble tube is a simple measuring device, but its use in LBE must be carefully assessed. Therefore, preliminary tests of bubble tubes in representative conditions have been carried out in a smaller test section. Experimental tests were performed at several temperatures, with LBE in stagnant conditions. The results obtained in these tests, aiming at checking the performance of the bubble tubes adopted in measuring pressure, differential pressure and level in the CIRCE facility, are discussed here. The obtained information will allow to calibrate the related measuring systems and to verify the accuracy and repeatability of the measurements, as a function of the injected gas flow rate, the tube diameter and the geometry of the tube exit section.     

Numerical analysis of bubble motion with the VOF method

Numerical analyses of a two-dimensional single bubble in a stagnant liquid and in a linear shear flow were conducted in the present study using the volume of fluid method, which is based on the local-instantaneous field equations. It was clarified that this method gives qualitatively appropriate predictions for the effects of the Morton number and the Eotvos number on fluctuating bubble motion in a stagnant liquid. Calculated velocity and pressure distributions indicated that the Karman vortex causes a sinuous movement of the bubble. As for the bubble motion in a linear shear flow, calculated bubbles migrated in a lateral direction. The direction of the lateral migration agreed to available experimental data. It was also confirmed that (i) the direction or the magnitude of the lateral migration is affected by the Eotvos and the Morton numbers, and (ii) the interaction among the internal flow of the bubble, the wake of the bubble and the external shear flow plays an essential role for the lateral migration.     

An experimental investigation on penetration of buoyancy-driven cold liquid into a vertical channel with hot, forced-flow

An experimental investigation on penetration of cold liquid into a modeled subassembly channel under mixed convection, from the hot plenum of a prototypical fast breeder reactor (FBR) design has been conducted. The penetration phenomenon occurs under certain natural circulation conditions during the operation of the direct reactor auxiliary cooling system (DRACS) for decay heat removal and can influence the natural circulation head that determines the core flow rate and consequently the core cooling rate. In the present experiment, a simplified test section simulating the upper plenum and a subassembly channel was constructed wherein both temperature and velocity measurements were made to investigate the penetration of cold water into a vertical channel with upward flowing hot water. Upon comparing temperature and velocity data we found an overall similarity in their respective spatio-temporal distributions as expected. The data suggested correlations describing the onset of flow penetration and penetration depth into the channel, expressed as a combination of Reynolds and Grashof numbers for a given Prandtl number. Flow penetration was identified based on the temperature and velocity data. This work suggests a dimensionless grouping based on an order of magnitude analysis that satisfactorily casts the experimental data while noting the subtle differences with other investigations.     

Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal–hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.