Effect of inlet geometry on hot-leg U-bend two-phase natural circulation in a loop with a large diameter pipe

In relation to nuclear reactor accident and safety studies, experiments on hot-leg U-bend two-phase natural circulation in a loop with a relatively large diameter pipe (10.2 cm ID) was performed for understanding the two-phase natural circulation and flow termination during a small break loss of coolant accident in LWR. The loop design was based on the scaling criteria developed under this program and a horizontal section was inserted between the gas injector and the hot leg in order to investigate the effect of the vapor phase inlet section on the flow regimes and flow interruption. The loop was operated either in a natural circulation mode or in a forced circulation mode using nitrogen gas and water. Various tests were carried out to establish the basic mechanism of the flow termination as well as to obtain essential information on scale effects of various parameters such as the loop frictional resistance, thermal center, and pipe diameter. The void distribution in a hot leg, flow regime and natural circulation rate were measured in detail for various conditions. The termination of the natural circulation occurred when there was insufficient hydrostatic head in the downcomer side. The superficial gas velocity at the flow termination could be predicted well by the simple model derived from a force balance between the frictional pressure drop along the loop and the hydrostatic head difference. The bubbly-to-slug flow transition was found to be dependent on axial locations. It turned out that the inlet geometry affected the flow regime at the inlet of the hot leg, namely the void distribution in the hot leg.     

Two-group interfacial area transport equations at bubbly-to-slug flow transition

In relation to nuclear reactor accident and safety studies, experiments on hot-leg U-bend two-phase natural circulation in a loop with a relatively large diameter pipe (10.2 cm inner diameter) were performed for understanding the two-phase natural circulation and flow termination during a small break loss of coolant accident in LWRs. The loop design was based on the scaling criteria developed under this program and the loop was operated either in a natural circulation mode or in a forced circulation mode using nitrogen gas and water. Various tests were carried out to establish the basic mechanism of the flow termination as well as to obtain essential information on scale effects of various parameters such as the loop frictional resistance, thermal center and pipe diameter. The void distribution in a hot-leg, flow regime and natural circulation rate were measured in detail for various conditions. The termination of the natural circulation occurred when there was insufficient hydrostatic head in the downcomer side. The superficial gas velocity at the flow termination could be predicted well by the simple model derived from a force balance between the frictional pressure drop along the loop and the hydrostatic head difference. The bubbly-to-slug flow transition was found to be dependent on axial locations. It turned out that the formation of cap bubbles in the large diameter pipe caused the increased drift velocity, which would affect the prediction of the void fraction in the hot leg.     

Characteristics of two-phase natural circulation in freon-113 boiling loop

A freon-113 flow visualization loop for simulating the hot-leg U-bend natural circulation flow has been constructed and hot-leg two-phase flow behavior has been studied experimentally. From the present experiments, an understanding of the basic mechanisms of the two-phase natural circulation and flow termination were obtained. The power input, loop friction and the liquid level in the simulated steam generator played key roles in the overall flow behavior. Experimental results show that the flow behavior strongly depends on phase changes and coupling between hydrodynamic and heat transfer phenomena. Non-equilibrium phase-change phenomena such as flashing create unstable hydrodynamic conditions which lead to cyclic or oscillatory flow behaviors.     

Natural circulation experiments with a VVER reactor geometry

Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after Soviet-designed VVER pressurized water reactors. Using stepwise inventory reduction and small-break experiments, primary loop flow behaviour was studied over a range of coolant inventories. The tests revealed a trend toward decreasing primary side mass flow rate with inventory. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs and coolant flow into the hot legs changed from single to two-phase flow. The cause of this flow interruption was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. Finally, an experiment was conducted to demonstrate how loop seal refilling behaviour at low coolant inventories depends upon the steam flow rate through an individual hot leg. It was shown that loop seal refilling results when low steam velocities permit countercurrent flow in the upflow side of the loop seal.     

Method of Estimating Liquid Velocity in Hot Leg during Reflux Cooling of Natural Circulation in Pressurized Water Reactor

Natural circulation in a PWR has considerable attention since the TMI-2 accident as an alternative cooling method or recovery technique from certain kinds of accidents or transients involving a loss of pumped circulation. Among the three modes of natural circulation (i.e. single-phase, two-phase and reflux cooling), reflux cooling has not been well investigated in a PWR configuration. The present study was thus focused on reflux cooling of natural circulation and analytical method was developed to estimate the liquid velocity of the condensed liquid in a hot leg of a PWR.

The results of the present study showed that the liquid velocity and the liquid thickness are estimated as 2.7 m/s and 3.0 cm, respectively, at the hot leg inlet from the upper plenum for the typical PWR reflux condition (2% core power at 6.9 MPa). Therefore it was concluded that a flow-blockage of the steam flow from the core by the condensed liquid flow is unlikely to occur in a hot leg. The results are also useful for designing a special instrumentation for measuring the condensed liquid flow rate and the liquid thickness in an experimental test facility for reflux cooling test.     

Two-phase flow in a full-scale loop seal facility

An experimental program has been carried out to study two-phase behaviour of a PWR cold leg loop seal during loss-of-coolant accidents. The experimental facility comprises a full-scale cold leg with a reactor coolant pump simulator. Three separate air/water test series were performed to determine the onset of slugging in the horizontal pipe, the residual water mass and the total two-phase pressure drop in the loop seal.The results of flow regime transition experiments have been compared with smaller-scale experiments and with theoretical predictions to evaluate scaling criteria. The strong hysteresis of transitions found between the stratified and slug flow regimes depends on the loop seal geometry and U-tube oscillations.     

低干度自然循环流量漂移的特征曲线图谱分析

在5MW低温核供热堆全模拟试验回路(HRTL-5)上,实验观察到了低干度自然循环条件下的流量漂移现象.通过一个考虑了加热段欠热沸腾、上升段冷凝、闪蒸等物理过程的两相流动数学模型,编制了相应的计算程序,获得了自然循环特征曲线图谱及其运行曲线,确定了自然循环分岔图和静态不稳定边界图,进而提出了通过自然循环特征曲线图谱研究流量漂移的分析方法.分析表明:特征曲线图谱方法是研究自然循环静态不稳定的有效手段.增大系统压力、减小热流密度、增加入口单相阻力、减小出口两相阻力有利于避免自然循环流量漂移的发生.     

Natural circulation flow behavior at reduced inventory in a VVER geometry

This paper deals with the natural circulation flow characteristics of the VVER-440 geometry at reduced coolant inventory. Special emphasis is on the flow rate of the primary circuits during the two-phase flow regime. For studying two-phase natural circulation flow phenomena in a VVER geometry a series of cold leg small break loss-of-coolant accident (SBLOCA) tests was carried out in the PArallel Channel TEst Loop (PACTEL), a 1/305 volumetrically scaled model of a VVER-440 reactor. The tests were conducted with break areas ranging from 0.1 to 1.5 % of the scaled cold leg cross-sectional area of the reference reactor. A partial failure of the high-pressure injection system (HPIS) was assumed. The tests reveal a trend towards an increasing primary circuit mass flow rate with decreasing inventory. This contradicts the findings of earlier tests in multi-loop VVER geometry. With single-loop facilities, increased mass flow rates at reduced inventories have been reported before. The increase of the two-phase flow rate turns out to be a consequence of the combined effect of break size, pressure range and secondary side feed and bleed procedure. The physical phenomena of flow stagnation in the primary circuits, system pressurization, asymmetric loop flows, and loop seal clearing and refilling take place during the natural circulation cooling process from single-phase into two-phase and boiler–condenser modes. In addition, flow reversal in the undermost tubes of the horizontal steam generators (SG) is observed. These phenomena are discussed briefly while a general insight into the course of the tests is presented.     

Flow excursion phenomenon and its mechanism in natural circulation

An experiment was performed on the test loop (HRTL-5), which simulates the geometry and system design of a 5-MW nuclear heating reactor. In a wide range of inlet subcoolings, different flow modes, such as single-phase stable flow, subcooled boiling stable flow, subcooled boiling static flow excursion, density-wave oscillation and stable two-phase flow in the natural circulation system have been described. The phenomenon and mechanism of the static flow-excursion, which has never been studied well on this field, is especially interpreted. The experimental results show that, in the process of flow excursion, the mass flow rate and the inlet temperature decreases, while the exit temperature increases smoothly. As the process of the excursion continues for about 1 h, short period dynamic flow oscillation occurs, which can only be seen in the process of this static flow excursion, and has also never been studied well. These static and dynamic flow instabilities combine together and continue for about 2 h, then a point is reached, at which the static flow excursion disappears, but the dynamic flow oscillation continues. The mechanism of the static flow excursion is interpreted through two sets of curves for flow resistance pressure drop and driven head in natural circulation, and one curve for the natural circulation operation under special thermohydraulic condition. The study of the flow excursion and its concerned dynamic flow oscillation is of great significance for the development of the nuclear heating reactor under natural circulation.     

立式倒U型管蒸汽发生器倒流现象及初步分析

文章涉及中国核动力研究设计院自然循环实验装置单相稳态自然循环实验过程中立式倒U型管蒸汽发生器(UTSG)模拟体一次侧流体的流动特性。实验观察到：1)UTSG模拟体进口腔室压力低于出口腔室压力；2)UTSG模拟体入口腔室温度较热段温度有一陡降。通过对该实验现象的分析可以判定，在单相自然循环工况下，UTSG模拟体中某些传热管内出现了倒流。实验结果表明，倒流的出现使UTSG模拟体自然循环工况下的流动阻力系数较强迫循环工况下的明显增大。      

Assessment of Some Cross Section Libraries through Analyses of Coupled-Core Critical Experiments by Monte Carlo Calculations

An experiment was performed on the natural circulation test loop HRTL-5, which simulates the geometry and system design of the 5 MW full power natural circulation nuclear heating reactor. Different flow modes, including density wave oscillation and flow excursion et al., were observed in a wide range of inlet sub-cooling at 1.5MPa. By means of self-developed computational codes, the bifurcation chart has been obtained. Consequently the flow excursion boundary has been determined. Through the analysis on the excursion boundary, the method to avoid the flow excursion during startup has been presented. Analytical results show: (1) with the decreasing heat flux or the increasing system pressure, the static flow excursion occurs at higher inlet temperature and its range in the instability maps becomes narrower correspondingly; (2) to decrease the outlet two-phase resistance or increase the inlet single-phase resistance is beneficial to avoid the flow excursion; (3) by means of increasing the system pressure to start up the reactor with low heat flux, the flow excursion and low steam quality density wave oscillation can be successfully avoided. This investigation is meaningful to the reactor safety and the design of the nuclear heating reactors.     

Characteristics of Type-I Density Wave Oscillations in a Natural Circulation BWR at Relatively High Pressure

Experiments were conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney at high pressure. The SIRIUS-N facility was designed to have non-dimensional values which are nearly equal to those of a typical natural circulation BWR. The observed oscillations are found to be density wave oscillations, since the void fractions in the chimney inlet and exit are out of phase. They belong to the Type-I category, since they occur at low flow qualities, according to the Fukuda—Kobori's classification. Moreover, the oscillation period correlates well with the passing time of bubbles in the chimney section regardless of the system pressure, the heat flux, and the inlet subcooling. Two distinct phenomena are found in relation between the oscillation period and liquid passing time in the chimney, indicating that the driving mechanisms of the instabilities are different between low and high pressures. Stability maps were obtained in reference to the inlet subcooling and the heat flux at the system pressures of 1, 2, 4, and 7.2 MPa. The flow became stable below a certain heat flux regardless of the channel inlet subcooling. The stable region enlarges with increasing system pressure. Thus, the stability margin becomes larger in a startup process of a reactor by pressurizing the reactor sufficiently before withdrawing the control rods. The obtained stability map demonstrates that the nominal operating condition of the ESBWR has a significant stability margin to the unstable region.     

Linear Analysis of Thermo-hydraulic Instabilities of the Advanced Heavy Water Reactor (AHWR)

Analysis was carried out to predict the threshold of instability for Ledinegg type and density wave oscillations for the Indian Advanced Heavy Water Reactor (AHWR) which is a Natural Circulation Pressure Tube Type Boiling Water Reactor. The mathematical model considers homogeneous two-phase flow and the conservation equations are solved analytically to obtain the steady state thermo-hydraulic characteristics and flow stability map. The model was applied for the AHWR concept after it had been validated with the test data obtained from a simple forced circulation loop with small parallel boiling channels and from the High Temperature Loop (PNC). The results indicate that the proposed design configuration of the AHWR may experience both Ledinegg type (static instability) and Type-I and Type-II density wave oscillations depending on the operating condition. The effects of various geometric and operating parameters on these types of instabilities were studied. It can be seen from the results that the Ledinegg type instability is suppressed with an increase in pressure and disappears when the operating pressure is higher than 0.7MPa. But the density wave instability may occur even at 7 MPa. In addition, it is found that for parallel multiple channels operating under natural circulation condition, the stability of density wave instability is not always enhanced by increasing the throttling coefficients at the inlet of channels.     

Analysis of the OECD/NEA ROSA Project experiment simulating a PWR small break LOCA with high-power natural circulation

OECD/NEA ROSA Project experiment with the large scale test facility (LSTF) in JAEA was conducted simulating a PWR 1% cold leg small break LOCA with an assumption of high-power natural circulation due to failure of scram and total failure of high pressure injection system. The core power curve for the LSTF experiment was obtained by PWR LOCA analysis using JAEA-developed coupled three-dimensional kinetics and thermal-hydraulics code SKETCH-INS/TRAC-PF1 with detailed core model. A post-test analysis was performed against the obtained data by using JAEA-modified RELAP5/MOD3.2.1.2 code to validate the code predictability. The JAEA-modified RELAP5/MOD3.2.1.2 code was used by incorporating a break model that employs maximum bounding flow theory with a discharge coefficient of 0.61 for two-phase break flow. In the experiment, flow in hot legs became supercritical during two-phase natural circulation, causing the hot leg liquid level to be quite low. Liquid accumulation in steam generator U-tube upflow-side took place during reflux condensation mode due to high vapor velocity. The RELAP5 code predicted reasonably well the overall thermal-hydraulic phenomena observed in the experiment. The code, however, overpredicted the break flow rate especially during two-phase flow discharge period probably because of the failure in the correct simulation of the cold leg liquid level due to late decrease in the primary loop flow rate.     

低压自然循环系统流动闪蒸过程流型研究

本文采用高速摄影和网格电导传感器对低压自然循环系统垂直上升段内闪蒸诱发的两相流流型演变开展研究。针对不同的流动状态，分别给出了稳定和不稳定两相流动条件下上升段内的流型种类。基于上升段内流体温度沿轴向的变化规律，确定流体温度沿轴向位置的转折点为闪蒸发生的起始位置。采用无量纲过冷数和闪蒸数，对低压自然循环系统的流动状态进行了划分；在入口过冷数小于12、闪蒸数介于4～5之间时，系统处于稳定的两相自然循环流动状态。     

Study on Pb-Bi natural circulation phenomena

For the development of 45w%Pb-55w%Bi cooled direct contact boiling water small fast reactor (PBWFR), experimental study on Pb-Bi-water direct contact boiling two-phase flow has been performed using Pb-Bi-water direct contact boiling two-phase flow loop. For stable start-up of the boiling flow operation, Pb-Bi single-phase natural circulation must be realized in a Pb-Bi flow system of the loop before water injection into Pb-Bi. The Pb-Bi flow system consists of a four-heater-pin bundle, a chimney, an upper plenum, a level meter tank, a cooler, and an electromagnetic flow meter. A stable Pb-Bi single-phase natural circulation was realized in the range of flow rate from 1.5 l/min to 4.8 l/min by heating Pb-Bi in the heater-pin bundle with a power up to 7.7 kW. The inlet and outlet temperatures of the heater bundle were in the ranges from 243°C to 278°C, and from 251°C to 278°C, respectively. The natural circulation flow was simulated analytically using one-dimensional flow model including frictional, form and drag forces. Total hydraulic head through the loop were calculated from Pb-Bi densities at measured Pb-Bi temperatures in the loop. It was found that the calculated flow rate agreed well with the measured ones, which indicated the validity of the analytical models.     

用集总参数法分析低含汽量自然循环回路汽液两相流稳定性

采用集总参数法分析低含汽量自然循环回路汽液两相流稳定性。描述热工水力现象的系统方程由均流模型偏微分守恒方程经集总参数平均导出，高含汽量常微分方程解程序包ＬＳＯＤＥ被用来解以常微分方程表征的系统方程，与清华大学核能技术研究院为分析５ＭＷ低温核供热堆热工水力特性而设计运行的两相流稳定性实验结果比较表明，采用集总参数法分析低含的自然循环回路两相流密度波振荡及其有关非线性现象是有效的。     

Sub- to Supercritical Flow Transition in a Horizontally-Stratified Two-Phase Flow in PWR Hot Legs

A co-current, horizontally-stratified, two-phase flow would appear in the hot legs of a pressurized water reactor during a certain class of small-break loss-of-coolant accident. The liquid velocity in the hot leg may become so high that it exceeds the speed of interfacial waves. The condition for the onset of such a “supercritical” flow is studied in this paper by analyzing experimental data taken in the ROSA-IV Large Scale Test Facility. It is shown that the energy loss at the hot leg inlet needs to be taken into account to predict the above onset condition reasonably well.     

Non-dimensional analysis of static bifurcation in a natural circulation loop

The steady-state characteristics of a two-phase natural circulation loop were investigated based on the homogenous model. Transcendental equations of non-dimensional loop mass flow rate under various conditions were also derived. The static bifurcation diagram of a two-phase natural circulation described with non-dimensional variables Npch-m^＋ was obtained. In addition, various steady-state characteristics of a natural circulation loop were analyzed and discussed. These characteristics include the existence of multiple solutions under certain conditions, and the maximum mass flow rate. The authors also examined the effects of important parameters such as sub-cooling number, riser-to-heated-region length ratio, and riser-to-heated-region diameter ratio.     

Investigation on two-phase critical flow for loss-of-coolant accident of pressurized water reactor

The previous investigations were mainly conducted under the condition of low pressure,however,the steam-water specific volume and the interphase evaporation rate in high pressure are much different from those in low pressure,Therefore,the new experimental and theoretical investigation are performed in Xi‘an Jiaotong University.The investigation results could be directly applied to the analysis of loss-of -coolant accident for pressurized water reacor.The system transition characteristics of cold leg and hot leg break loss-of -coolant tests are described for convective circulation test loop.Two types of loss-of-coolant accident are identified for :hot leg” break,while three types for “cold leg”break and the effect parameters on the break geometries.Tests indicate that the mass flow rate with convergent-divergent nozzle reaches the maximum value among the different break sections at the same inlet fluid condition because the fluid separation does not occur.A wall surface cavity nucleation model is developed for prediction of the critical mass flow rate with water flowing in convergentdivergent nozzles.