Thermal hydraulic performance analysis of a double tube bundle steam generator for a liquid metal reactor

In this study, thermal-hydraulic performance of a double tube bundle steam generator (DTBSG) using helically coiled tubes was analyzed numerically. For this purpose a one-dimensional thermal-hydraulic analysis computer program was developed.     

CPR1000蒸汽发生器二次侧三维稳态热工水力分析

采用三维稳态分析软件GENEPI,对CPR1000蒸汽发生器二次侧管束区进行了热工水力计算,利用多孔介质及局部阻力系数来表征传热管及各几何部件的复杂结构和压降影响,得到了二次侧管束区流场、温度场等的分布情况。计算结果表明:管束区最大干度为0.3;将典型传热管的动能数据提供给流致振动软件进行计算分析,结果显示在本工况下,传热管的流致振动在可接受范围内;对管板附近的流场及温度场进行分析,预测了此模型及工况下的泥渣沉积区域,为排污管的设计提供了输入数据。计算结果验证了CPR1000蒸汽发生器二次侧管束区设计的合理性。     

小型模块式反应堆螺旋管蒸汽发生器设计和热工水力分析

近年来,国际上一体化小型模块式反应堆发展飞速,我国也正在加速研制一体化小型模块式反应堆。本文针对15 MW的一体化小型模块式反应堆,设计一种螺旋管式蒸汽发生器,共12个蒸汽发生器组件均匀分布在反应堆堆芯围板外侧和压力容器内侧壁的环形空间中,每个组件含5层、25根螺旋管,整个蒸汽发生器共300根螺旋管。给出了蒸汽发生器的具体参数,分析了蒸汽发生器组件中换热系数、温度、温差和热流密度等沿管长的变化,并给出了螺旋管内流体的动力特性曲线。     

Dynamic characteristics of steam generator U-tubes with defect

This study investigates the fluid elastic instability characteristics of steam generator (SG) U-tubes with defect and the safety assessment of the potential for fretting-wear damages caused by foreign object in operating nuclear power plants. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for the U-tubes either with axial or circumferential flaw with different sizes. Special emphases are on the effects of flaw orientation and size on the modal and instability characteristics of tubes, which are expressed in terms of the natural frequency, corresponding mode shape and stability ratio. Also, the wear rate of U-tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the flow velocity and vibration of the tube on the remaining life of the tube. In addition, addressed in this study is the effect of the internal pressure on the vibration and fretting-wear characteristics of the tube.     

基于RELAP5的螺旋管蒸汽发生器热工水力程序研发与验证

由于较高的换热效率和紧凑的结构设计,螺旋管式直流蒸汽发生器（HCOTSG）在多种模块化小型堆的设计中得到了广泛应用。RELAP5作为广泛应用于反应堆热工水力特性分析的大型系统程序之一,采用的热工水力关系式仅针对直管模型开发,不适用于HCOTSG一次侧和二次侧。本文选用螺旋管及横掠管束的热工水力模型,基于RELAP5程序开发了HCOTSG模块。采用实验数据及程序对比等方式对螺旋管模块的流动和换热模型进行了单独验证,利用开发的RELAP5-HCOTSG程序针对国际革新安全反应堆（IRIS）的蒸汽发生器设计进行了整体的热工水力模拟,与原始RELAP5的计算相比,RELAP5-HCOTSG程序计算得到的热工水力参数与设计值符合良好,确认了本文开发的程序模块在HCOTSG热工水力分析中的适用性。     

高温气冷堆螺旋管蒸汽发生器流量漂移不稳定性研究

建立与氦气对流换热的并联螺旋管蒸汽发生器数值模型,分别采用一维飘移流模型和一维可压缩流动模型描述水侧和氦气侧的流动。在此基础上研究了球床模块式高温气冷堆核电站螺旋管蒸汽发生器内的流量漂移不稳定性。动态计算结果表明,在一定条件下蒸汽发生器内有可能发生流量漂移,不同传热管流量可相差几倍,而出口温度则相差几百度。通过对质量流速-压降曲线的分析,发现热负荷对稳定性起主导作用,热负荷越大越易发生流量漂移,且边界质量流量随热负荷呈线性增长。增大入口节流阻力和过冷度可以在一定程度上避免流量飘移。最后给出了蒸汽发生器流量飘移的稳定边界。     

影响管阵流体弹性不稳定性因素的研究

研究了管阵中管数,纵向流的作用以及U形管等对管阵流体弹性不稳定性的影响。介绍了用求解模态展开后的振动方程计算管阵的临界流速的方法。给出了简化处理耦合问题的单一振型法和频率截断法。并推导出考虑非均勻流、纵向流存在和U形管的基本方程。     

Friction Pressure Drop and Heat Transfer Coefficient of Two-Phase Flow in Helically Coiled Tube Once-Through Steam Generator for Integrated Type Marine Water Reactor

Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report.     

Flow-induced vibration: recent findings and open questions

Component failures due to excessive flow-induced vibration are still affecting the performance and reliability of nuclear power stations. Tube failures due to fretting-wear in nuclear steam generators, and vibration related damage of reactor internals are of particular concern. The purpose of this paper is to review some of the recent findings in the area of flow-induced vibration and to discuss some of the remaining questions. Vibration excitation mechanisms and damping mechanisms are described with particular emphasis on fluidelastic instability and damping in two-phase flows. The need for a better understanding of two-phase flow regimes, particularly in cross flow, is outlined. The dynamic characteristics of nuclear structures are explained. The statistical nature of some parameters, in particular support conditions, is discussed. The prediction of fretting-wear damage is approached from several points of view. An energy approach to formulate fretting-wear damage is proposed.     

Vibration analysis of heat exchanger and steam generator designs

A thorough flow-induced vibration analysis of nuclear components such as heat exchangers and steam generators is essential at the design stage to ensure good performance and reliability. This paper presents our approach and techniques in this respect. In a steam generator, for example, the flow may be liquid or two-phase. In general, parallel and cross-flow exist in the tube bundles of heat exchange components. In cross-flow three basic vibration excitation mechanisms are considered, namely fluidelastic instability, periodic wake shedding resonance, and forced response to random flow turbulence. The latter may need to be considered in parallel flow. These vibration excitation mechanisms and the dynamics of multispan tubes are formulated in a computer model which is used to predict the vibration response of the tubes. The computer model and the parameters required to formulate the vibration excitation mechanism are discussed. Examples of vibration analysis of steam generators and heat exchangers are outlined. It is concluded that most flow-induced vibration problems may be avoided by proper analysis at the design stage.     

传热管与支撑件随机碰撞实验研究

针对蒸汽发生器中传热管与支撑件的碰撞行为,对悬臂梁固定的传热管在不同支撑条件下开展了激振实验,获得了传热管均方根位移与接触率,分析了传热管与支撑件磨损功率的变化规律,并探究了传热管固有频率对振动特性的影响。结果表明,防振条支撑与波纹带支撑时传热管的法向均方根位移均随激振力增加逐渐放缓,而防振条支撑对应的切向位移呈线性增长。防振条支撑与波纹带支撑时的接触率均表现为随激振力增大趋于稳定,其中间隙对防振条支撑的接触率影响更明显。在以冲击为主导的激励方式下,激振力与磨损功率表现为明显的正相关。支撑间隙对磨损功率的影响相对复杂,防振条支撑下磨损功率在0.1 mm和0.25 mm间隙存在极值,而波纹带支撑磨损功率仅在0.2 mm间隙存在极值。传热管固有频率对振动响应结果的影响很小。     

螺旋管核态沸腾两相流摩擦压降与传热数值分析

通过合理简化一体化模块式先进反应堆(SMART)螺旋管式蒸汽发生器,建立螺旋管单元管模型,采用两流体模型和非平衡过冷沸腾模型,在均匀热流密度下对螺旋管内流体进行不同参数下流动与传热数值模拟。结果表明：摩擦压降数值计算结果与陈学俊经验公式最为接近;曲率从0.04降至0.012时,摩擦压降明显下降,曲率继续下降,摩擦压降不变;加速压降几乎不受曲率影响;螺旋升角为3°～8.6°时,计算摩擦压降可不考虑螺旋升角的影响;雷诺数越大,总压降和摩擦压降均变大,摩擦压降梯度也明显增大。     

Heat transfer characteristics of horizontal steam generators under natural circulation conditions

This paper deals with the heat transfer characteristics of horizontal steam generators, particularly under natural circulation (decay heat removal) conditions on the primary side. Special emphasis is on the inherent features of horizontal steam generator behaviour. A mathematical model of the horizontal steam generator primary side is developed and qualitative results are obtained analytically. A computer code, called HSG, is developed to solve the model numerically, and its predictions are compared with experimental data. The code is employed to obtain for VVER 440 steam generators quantitative results concerning the dependence of primary-to-secondary heat transfer efficiency on the primary side flow rate, temperature and secondary level. It turns out that the depletion of the secondary inventory leads to an inherent limitation of the decay energy removal in VVER steam generators. The limitation arises as a consequence of the steam generator tube bundle geometry. As an example, it is shown that the grace period associated with pressurizer safety valve opening during a station black-out is hours instead of the 5–6 hours reported in several earlier studies. (However, the change in core heat-up timing is much less—about 1 h at most.) The heat transfer limitation explains the fact that, in the Greifswald VVER 440 station black-out accident in 1975, the steam generators never boiled dry. In addition, the stability of single-phase natural circulation is discussed and insights on the modelling of horizontal steam generators with general-purpose thermal-hydraulic system codes are also presented.     

压水堆蒸汽发生器一、二次侧稳态流场耦合分析

蒸汽发生器（SG）在运行过程中主要面临流致振动所导致的传热管破裂事故,而流致振动分析需以SG内的三维两相流场作为输入条件。采用多孔介质模型,对SG二次侧流场进行求解,同时耦合一、二次侧换热,获得SG二次侧速度场、温度场、压力场及流动含气率分布,并获得传热管一维的一、二次侧流体温度和换热系数及传热管温度分布。由于一次侧向二次侧释热极不均匀,SG内流场分布及汽水分离器内的含气率分布极不均匀;汽水分离器内的最大、最小含气率分别为0.62和0.05,该参数可为汽水分离器负载设计提供依据。通过计算还获得弯管区速度分布,该分布可为传热管的流致振动磨损评估提供输入条件。     

蒸汽发生器管束的固有振动分析

通过对蒸汽发生器管束模型的建立,采用有限元分析方法,研究了支撑板、流量分配板、防振杆、环箍等部件对管束固有振动的影响.同时考虑了传热管内外流体对管束固有频率的影响.计算表明:防振杆与环箍的结构和尺寸对蒸汽发生器管束的固有振动频率有显著影响.     

蒸汽发生器传热管束过冷沸腾区两相流动数值模拟

基于两流体欧拉数学模型结合RPI壁面沸腾模型,利用大型商用CFD软件ANSYS CFX 12.0对蒸汽发生器传热管束过冷沸腾区一次侧、壁面和二次侧耦合传热过程进行了数值模拟。研究了三叶梅花孔支撑板和不同入口过冷度条件下蒸汽发生器传热管束内的流动沸腾现象,得到一、二次侧流场与温度场,二次侧空泡份额分布,支撑板梅花孔局部的流动状况及不同入口过冷度对蒸汽发生器热工水力特性的影响。数值模拟结果表明,三叶梅花孔支撑板的存在及不同入口过冷度对蒸汽发生器传热管束过冷沸腾区域的热工水力特性影响显著。     

Modelling of fluidelastic vibrations of heat exchanger tubes with loose supports

The existence of gaps at tube supports necessitates time domain modelling of fluid forces to predict flow-induced vibrations and associated wear in heat exchangers and steam generators. This paper presents a new time-domain model for fluidelastic instability forces of tubes with loose-supports. In this model, the fluidelastic force, which is dependent on flow velocity and array geometry, is superimposed on the turbulence forcing function. The model was used to calculate the critical flow velocity, tube response, and tube/support interaction parameters, such as impact force and work rate. The critical velocity for linear cases was accurately predicted. The critical flow velocity for the loose support case was found to be sensitive to both the gap size and the turbulence level.     

The effect of tube rupture location on the consequences of multiple steam generator tube rupture event

A multiple steam generator tube rupture (MSGTR) event has never occurred in the commercial operation of nuclear reactors while single steam generator tube rupture (SGTR) events are reported to occur every 2 years. As there has been no occurrence of a MSGTR event, the understanding of transients and consequences of this event is very limited. In this study, a postulated MSGTR event in an advanced power reactor 1400 (APR1400) is analyzed using the thermal-hydraulic system code, MARS1.4. The APR1400 is a two-loop, 3893 MWt, PWR proposed to be built in 2010. The present study aims to understand the effects of rupture location in heat transfer tubes following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 allows the shortest time for operator action following a tube rupture in the vicinity of the hot-leg side tube sheet and allows the longest time following a tube rupture at the tube top. The MSSV lift time for rupture at the tube-top is evaluated as 24.5% larger than that for rupture at the hot-leg side tube sheet.     

双弹性管流固耦合振动的数值模拟

为研究反应堆结构中诸如燃料棒、蒸汽发生器和其他换热器等管束类结构的流固耦合振动问题,利用有限体积法离散大涡模拟的流体控制方程及有限元方法离散结构动力学方程,结合动网格技术,建立了三维流体诱发弹性管束振动的数值模型,实现了计算结构动力学与计算流体力学之间的双向耦合。得到横流作用下单管的振动响应,并与已有的实验数据比较,证明了本文模型的合理性;对横流作用下的两串列管、两并列管的流固耦合振动进行了数值模拟,着重研究了节径比为1.2、1.6、2、3、4的两弹性管在不同流速作用下的动力学响应及流场特性;得到串列管、并列管的临界间距与临界流速。     

Experimental validation of the helical steam generator model in the TASS/SMR code

The transients and setpoint simulation/system-integrated modular reactor (TASS/SMR) code has been used to identify the safety margin of a 65-MWt advanced integral reactor and to evaluate its design performance. Although, the code has been verified by using simplified and analytical problems as well as a reliable system code, its validation has not been fully established. This paper deals with a validation of the TASS/SMR code by using two kinds of separate effect tests related to heat transfer at a helically coiled steam generator. The heat transfer experiments were performed by using a full-scale prototype of the steam generator cassette of the advanced integral reactor and a scaled-down steam generator cassette. Analytical results show that the TASS/SMR code predicts the thermal hydraulic parameters, including the system pressure and fluid temperature at the primary and secondary sides of the steam generator cassette, and the heat transfer rate through the steam generator cassette well. The validation results in this study show that the TASS/SMR code is applicable for heat transfer calculations related to the helically coiled steam generator of the advanced integral reactor.