Effect of pump-induced cold-leg swirls on the flow field in the RPV of the EPR™: CFD investigations and comparison with experimental results

A computational fluid dynamic (CFD) model for the pressure vessel of the evolutionary pressurized reactor (EPR™) was developed and validated. The aim of this model is the simulation of transients where three-dimensional effects play a strong role, such as boron dilution and main steam line break (MSLB) scenarios. First, a full solid (CAD) model has been built, that includes all details of the reactor pressure vessel (RPV) and the internals which are important for fluid dynamic analyses. The solid model has then been used as basis for the generation of the computational mesh necessary to carry out CFD simulations. Both a hexahedral and a polyhedral mesh have been created. The CFD model has been validated against experimental results of the JULIETTE facility, a 1:5 scaled mock-up of the EPR™ reactor RPV built by AREVA and equipped with advanced instrumentation.The performances of the hexahedral and the polyhedral meshes are investigated in relation to the agreement with experimental data, convergence and CPU requirements. In addition, the effect of the cold-leg swirls on the velocity field inside the RPV is investigated. These swirls mimic the effects of the main coolant recirculation pumps on the flow field at the entrance of the RPV. It is shown that the CFD model is able to capture the shift of the maximum velocity in the downcomer annulus observed in the experimental results. Good qualitative as well as quantitative agreement with the experimental data is achieved.     

ACP600主蒸汽管道破裂事故的应对措施研究

针对ACP600取消高压安注系统和浓硼箱、使用一体化钆为可燃毒物、采用Mode-C运行与控制模式等设计改进导致主蒸汽管道破裂（MSLB）事故安全裕量降低的不利情况,对先进三代核电厂ACP600的MSLB事故进行分析研究。为避免MSLB事故下反应堆重返临界后堆芯功率峰值过高导致偏离泡核沸腾比（DNBR）低于限制值,分别从快速注入硼溶液和减缓堆芯冷却率的角度,评价不同的安注系统配置以及停运故障环路主泵对于缓解MSLB事故的作用。研究最佳的缓解方案,并提出增设“蒸汽管线压力低-3”信号停运故障环路主泵的设计优化建议。      

Evaluation of Acoustic- and Flow-Induced Vibration of the BWR Main Steam Lines and Dryer

The boiling water reactor (BWR-3) steam dryer in the Quad Cities (QC) Unit 2 Nuclear Power Plant was damaged by high-cycle fatigue due to acoustic-induced vibration. The cause of the dryer failure was considered as flow-induced acoustic resonance at the stub pipes of the safety relief valve (SRV) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by the interaction between the sound field and an unstable shear layer across the closed side branches of SRVs. We have started a research program on BWR steam dryers to develop methods of evaluating the loading. Moreover, it is necessary to evaluate the dryer integrity of BWR-5 plants, which are the main type of BWR in Japan. In the present study, we conducted 1/10-scale BWR model tests and analysis to investigate the flowinduced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome, and 4 MSLs with 20 SRV stub pipes. Computational fluid dynamics (CFD) analysis was conducted to evaluate the acoustic source in MSLs. Finite element method (FEM) was applied to calculate the three-dimensional wave equations for acoustic analysis. We demonstrated that large fluctuating pressure occurred in the high- and low-frequency regions. The high-frequency fluctuating pressure was generated by the flow-induced acoustic resonance in the SRV stub pipes. We evaluated the acoustic source (that is, the fluctuating pressure) in MSLs by unsteady CFD calculations, and we evaluated the pressure propagation by acoustic analysis. These results were verified by comparison with the results of scale-model tests, and they showed good agreement with the experimental results. The effects of the difference between the properties of air and steam were numerically investigated, and it was found that the effects on the acoustic resonance in the SRV stub pipes were not significant.     

红沿河核电厂VVP/ARE隔室改进超压分析

本文对拟建于严寒地带的红沿河核电厂CPR1000机组中主蒸汽管道和主给水管道(VVP/ARE)隔室与外界环境之间的封堵进行了设计改造,并对改造后方案的隔室墙体重新进行由隔室内管道破裂引起的超压风险分析论证。采用隔室热工响应分析程序对不同封堵方案进行计算分析,对比分析了不同封堵方案下不同隔室发生主蒸汽管道双端剪切断裂（MSLB）事故后的超压后果,论证了封堵方案的可行性。文中还针对封堵方案进行了敏感性研究,并给出了最佳封堵方案。该封堵方案已在红沿河核电厂实施。     

Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.     

Modeling and optimization of steam dryer for removal of lead–bismuth droplets

Removal of lead–bismuth droplets from steam flow is a crucial issue in the direct contact boiling lead–bismuth cooled fast reactor. Droplets are generated due to the boiling of water directly in the reactor chimney, where steam for the turbine is generated. The droplets could severely damage the turbine and therefore a steam dryer is used for their removal. This paper presents an optimization of the main steam dryer geometrical parameters and steam inlet velocity. The Lagrangian method is used, in which first the steam flow field is developed using the CFD code FrontFlow/Red and then the particle motion is simulated. It was found that the reduction of the plate spacing can improve the steam dryer performance without a significant increase of pressure drop, the wane pitch has a value after which the steam dryer performance is not significantly improved, the number of wanes of 1.5 was selected at this point, however, a more detailed model is necessary to arrive at the final conclusion. The optimum steam inlet velocity should be found using a detailed economical assessment. Velocities between 2 and 4 m/s seem to be reasonable to achieve good removal efficiency and keep the pressure drop at reasonable values.     

RPV侧边破口条件下堆外蒸汽爆炸计算分析

针对实际过程中更有可能发生的压力容器（RPV）侧边破口条件开展蒸汽爆炸计算分析。根据经济合作与发展组织（OECD）发布的现象识别与重要度排序表（PIRT）,选取堆外蒸汽爆炸敏感性分析参数,使用MC3D软件建立三维局部破口和二维环状破口几何模型,对影响计算结果的重要参数（破口尺寸、堆坑水位、破口位置、触发条件、液柱碎化和液滴碎化模型）开展RPV侧边破口条件下敏感性分析,获得最恶劣计算工况条件。敏感性分析结果表明,在大破口失水事故（LBLOCA）工况下,当堆坑处于满水位、RPV发生二维侧边环状破口、接触堆坑侧壁面时触发蒸汽爆炸、采用CONST模型和Classical模型时,堆坑侧壁面的压力载荷计算结果最为保守,对堆坑和安全壳完整性威胁最大。      

秦山核电厂主蒸汽管道破裂事故的分析研究

文章给出了压水堆核电厂主蒸汽管道破裂事故(MSLB)的概述、分析模型及主要假设,讨论了秦山核电厂影响MSLB的参数特点,并给出了极限工况的分析结果及敏感性分析得到的结论。     

Fluid-structure interaction analysis of large-break loss of coolant accident

In this article, we study a Large-Break Loss of Coolant Accident (LBLOCA) where a guillotine break of one of the main coolant pipes occurs near the reactor pressure vessel (RPV). This initiates a rarefaction wave which propagates inside the RPV. The simulation of bidirectional fluid-structure interaction phenomena has been found important for accurate prediction of the resulting deformation and loads. In this article, fully coupled simulation results are validated against the German HDR (Heißdampfreaktor) experiments. The computational fluid dynamic (CFD) software Fluent and Star-CD are applied to modelling of three-dimensional, viscous, turbulent fluid flow. The MpCCI code is used for bidirectional coupling of the CFD simulation to the structural solver Abaqus. Pressure boundary condition at the pipe break is obtained in a two-phase simulation with the system code APROS. Comparisons are made for break mass flow, wall pressure, displacement and strain. The simulation results follow the experimental data fairly well. In addition, the sensitivity of the results to numerical methods, grid resolution and pressure boundary condition are studied following the Best Practice Guidelines.     

Analysis of maximum pressure in VVER1000/V446 reactor containment for LOCA and MSLB

The highest thermal-hydraulic pressure in the containment occurs when reactor coolant in the first loop and steam in the secondary loop discharge simultaneously,and when the maximum amount of energy from reactor unit enters to containment volume.In this paper,we investigate temperature and pressure variations in the VVER1000 containment compartments owing to concurrent break in the pipelines of the primary and secondary loops.A two-phase,multicellular model is applied in the presence of non-condensable gases.Convection and conduction through the main heat structures inside the containment are also considered.The predicted results agree well with available data.Maximum values of pressure and temperature in the containment are then calculated and compared to the design values.If LOCA and MSLB occur simultaneously,the maximum pressure would exceed the design value and integrity of the containment would be threatened.     

Numerical analysis and visualization experiment on behavior of borated water during MSLB with RCP running mode in an advanced reactor

The core bypass phenomenon of borated water injected through direct vessel injection (DVI) nozzles in APR1400 (Advanced Power Reactor 1400MWe) during main steam line break (MSLB) accidents with a reactor coolant pump (RCP) running mode has been simulated using a two-channel and one-dimensional system analysis model code (MARS), and a three-dimensional computational fluid dynamics (CFD) code (FLUENT). A visualization experiment has also been performed using a scaled-down model of the APR1400. The MARS analysis has predicted a serious core bypass phenomenon of borated water, while the CFD analysis has shown results opposite to the MARS results. The CFD analysis has shown that the flow pattern in the downcomer is fully three-dimensional and that vortex flow structures are formed near the cold legs so that the borated water might pass without difficulty into the high flow region of the cold legs and flow well into the lower downcomer. The visualization experiment has shown that the borated water flows well to the lower plenum, as in the CFD analysis. Both the CFD analysis and visualization experiment have proved that a serious core bypass phenomenon of borated water might not happen in the APR1400. These results are quite different from those predicted by MARS.     

Dynamic response of reactor internals to pipe breaks

This paper investigates the horizontal responses of the reactor internals to a 14 inch safety injection nozzle break, which is expected to cause the largest loads of the branch line pipe breaks defined for a nuclear power plant. It examines the effects of two forcing terms, reactor vessel motions and internal hydraulic loads, and suggests a new procedure which can be used for tributary pipe break analysis. The analysis result confirms the applicability of the suggested procedure to small size tributary pipe break analysis. Also, this paper calculates the horizontal responses of the reactor internals to a 3 inch pressurizer spray line nozzle break, which is the only one remaining on the primary side after leak-before-break evaluation, and secondary side pipe breaks such as main steam line and economizer feedwater line. The responses are compared with those of safe shutdown earthquake to show that seismic loads with a conservative margin may be used for the pipe break loads in the preliminary design.     

Air/steam flow and steam wetness dependence on acoustic resonance in safety relief valves

Many experimental studies related to the flow-induced acoustic resonance closed side branches have been reported. However, few studies have reported on the effects of air/steam flow and steam wetness dependence on fluctuating pressure amplitude. Therefore, we investigated the effect of air/steam flow and steam wetness dependence on fluctuating pressure amplitude by conducting a high temperature and high pressure tests at the Hitachi Utility Steam Test Leading Facility (HUSTLE). The test section consisted of a main pipe and a side branch. The side branch was mounted on the long straight main pipe. Fluctuating pressures at the end face of the side branches were measured. The following two results were obtained; the first is that the air/steam flow had little effect on the fluctuating pressure amplitude normalized by dynamic pressure and frequency normalized by the resonance frequency; the second is that under the acoustic resonance (St = 0.41) and non-resonance (St = 0.55) conditions, fluctuating pressure and frequency changed little with steam wetness. The steam wetness during the boiling water reactor operation was less than 0.1%; thus, there was no effect of steam wetness on the acoustic pressure amplitude and the frequency under this operating condition.     

A study on the minimum steam cooling pressure of EOP for BWR power plants

The minimum steam cooling pressure (MSCP) is an important parameter for safe operation of boiling water reactor (BWR)-type nuclear power plant for the anticipated transient without scram (ATWS) scenario with reactor pressure vessel (RPV) water level unknown. Under such situation, the operator is requested to open the safety/relief valves (SRVs) and control the RPV pressure slightly above the MSCP so that adequate core cooling can be maintained. It is derived based on steam cooling strategy.The MSCP, defined to be the lowest RPV pressure at which the covered portion of the core, is capable of generating sufficient steam to preclude peak cladding temperature (PCT) in the uncovered portion of the core from exceeding 1088 K (1500 °F). It is calculated by two parameters - (1) the minimum bundle steam flow (Wg-1500) to maintain PCT < 1088 K (1500 °F) and (2) the number of SRVs available for opening.For current emergency operating procedure (EOP), only one set of MSCP derived based on one value of Wg-1500 for the ATWS condition. Furthermore, it is derived based on decay power of 2.2% rated power. Thus, the current MSCP used for the ATWS accident scenarios was deemed inadequate. The purpose of this paper (work) is to study the MSCP used in the ATWS conditions. For case of ATWS of 13% full power, controlling RPV pressure at MSCP of current approach ends up with core melt. The Wg-1500 is suggested to be replaced by the steam generation rate at minimum steam cooling RPV water level (MSCRWL), which is a function of power level. Simulation result indicates controlling RPV pressure at MSCP is equivalent to controlling the RPV water level at MSCRWL. The revised MSCP is dependent on the ATWS power level.     

Critical crack depth diagram of reactor vessel for pressurized thermal shock

During the operation of a pressurized water reactor, a certain type of transients could induce rapid cooldown of the reactor pressure vessel (RPV) with relatively high or increasing system pressure. This induces a high tensile stress at the inner surface of the RPV, which is called the pressurized thermal shock (PTS). The structural integrity of the RPV during PTS should be evaluated assuming the existence of a flaw at the vessel. For the quantitative evaluation of the vessel failure risk associated with PTS, the probabilistic fracture mechanics (PFM) analysis technique has been widely used. But along with PFM analysis, deterministic analysis is also required to determine the critical time interval in the transient during which mitigating action can be effective. In this study, therefore, the procedure for the deterministic fracture mechanics analysis of RPV during PTS is investigated using the critical crack depth diagram and the computer program to generate it is developed. Four transients of typical PTS, steam generator tube rupture, small break loss of coolant accident and steam line break are analyzed, and their response characteristics such as critical crack depth and critical time interval from the initiation of the transient are investigated.     

Experimental study on the blowdown load during the steam generator feedwater line break accident in the evolutionary pressurized water reactor

Integral effect tests using the ATLAS facility were performed to obtain the thermal-hydraulic parameters such as dynamic and static pressures, local temperatures, and flow rates during a feedwater line break of a steam generator. The break of a feedwater line was simulated using a double rupture disc assembly in order to satisfy the requirements for the break opening time of around a few milliseconds. In the present study, estimated break opening time was less than 1.5 ms and broken areas were 48.1% and 93.4% of the feedwater line, respectively. The maximum dynamic pressures of about 1.57 bar were obtained inside of feedwater box that was closest to the break location of the feedwater line. After the break of the feedwater line, propagation of the pressure wave along the distance from the break location inside the steam generator was clearly and pertinently observed in all the tests. From a structural integrity point of view, however, the risk induced by this maximum dynamic load could be treated to be insignificant.     

严重事故条件下安全壳响应模拟研究

压水堆核电厂发生严重事故期间,从主系统释放的蒸汽、氢气以及下封头失效后进入安全壳的堆芯熔融物均对安全壳的完整性构成威胁。以国内典型二代加压水堆为研究对象,采用MAAP程序进行安全壳响应分析。选取了两种典型的严重事故序列：热管段中破口叠加设备冷却水失效和再循环高压安注失效,堆芯因冷却不足升温熔化导致压力容器失效,熔融物与混凝土发生反应（MCCI）,安全壳超压失效;冷管段大破口叠加再循环失效,安全壳内蒸汽不断聚集,发生超压失效。通过对两种事故工况的分析,证实了再循环高压安注、安全壳喷淋这两种缓解措施对保证安全壳完整性的重要作用。     

Steam leak detection in advance reactors via acoustics method

Prediction of LOCA (loss of coolant activity) plays very important role in safety of nuclear reactor. Coolant is responsible for heat transfer from fuel bundles. Loss of coolant is an accidental situation which requires immediate shut down of reactor. Fall in system pressure during LOCA is the trip parameter used for initiating automatic reactor shut down. However, in primary heat transport system operating in two phase regimes, detection of small break LOCA is not simple. Due to very slow leak rates, time for the fall of pressure is significantly slow. From reactor safety point of view, it is extremely important to find reliable and effective alternative for detecting slow pressure drop in case of small break LOCA. One such technique is the acoustic signal caused by LOCA in small breaks. In boiling water reactors whose primary heat transport is to be driven by natural circulation, small break LOCA detection is important. For prompt action on post small break LOCA, steam leak detection system is developed to detect any leak inside the reactor vault. The detection technique is reliable and plays a very important role in ensuring safety of the reactor. Methodology developed for steam leak detection is discussed in present paper. The methods to locate the leak is also developed and discussed in present paper which is based on analysis of the signal.     

Commissioning of the ATLAS thermal-hydraulic integral test facility

KAERI recently constructed a new thermal-hydraulic integral test facility for advanced pressurized water reactors (PWRs) – ATLAS. The ATLAS facility has the following characteristics: (a) 1/2-height&length, 1/288-volume, and full pressure simulation of APR1400, (b) maintaining a geometrical similarity with APR1400 including 2(hot legs) × 4(cold legs) reactor coolant loops, direct vessel injection (DVI) of emergency core cooling water, integrated annular downcomer, etc., (c) incorporation of specific design characteristics of OPR1000 such as cold leg injection and low-pressure safety injection pumps, (d) maximum 10% of the scaled nominal core power. The ATLAS will mainly be used to simulate various accident and transient scenarios for evolutionary PWRs, OPR1000 and APR1400: the simulation capability of broad scenarios including the reflood phase of a large-break loss-of-coolant accident (LOCA), small-break LOCA scenarios including DVI line breaks, a steam generator tube rupture, a main steam line break, a feed line break, a mid-loop operation, etc. The ATLAS is now in operation after an extensive series of commissioning tests in 2006.     

BWR steam dryer for extended power uprate

A new steam dryer for extended power uprated conditions, based on a design with proven performance record, has been designed by Westinghouse with performance characteristics meeting all utility requirements. Extensive use of computational fluid dynamics (CFDs) has been made to design the new dryer intended for plants of the BWR 3000-type. In addition to numerical simulations, the analyses rely on previous knowledge, acquired from experimental work on steam flow characteristics in the BWR 3000 with its asymmetrically placed steam line nozzles. The new design is expected to both decrease vibration levels in the steam lines and to solve a water-level measurement problem. An extensive experimental verification of the new design is currently in progress.