共查询到19条相似文献,搜索用时 234 毫秒
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在失水事故(LOCA)工况下安注系统投入使用时,蒸汽与安注冷却剂会发生流体热力学混合,热混合过程中冷腿段的冷却是直接影响堆芯再淹没与否的重要因素。中国广核集团有限公司自主研发了一款两相流热工水力系统分析软件LOCUST,可用于压水堆核电厂事故工况的分析计算。基于西安交通大学堆芯应急冷却系统(ECCS-XJTU)试验台架进行的堆芯应急冷却(ECC)安注热混合试验,本文使用LOCUST软件对ECC热混合试验进行了几何建模及计算分析。ECC热混合试验工况主要为不同流量下主管纯蒸汽与安注管过冷水的混合,蒸汽流量为25~125 kg/h,过冷水流量为100~500 kg/h。模拟计算结果和试验结果的对比分析表明:试验段出口质量流量计算值的最大相对误差在13.8%以内,混合后温度计算值的最大相对误差在8%以内,LOCUST在计算高温蒸汽和过冷水混合时的计算结果相对保守,总体上验证了LOCUST在LOCA下两相热混合安注计算的可靠性和准确性。 相似文献
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反应堆一回路系统在自然循环条件下,蒸汽发生器(SG)部分U型管内可能会出现回流现象,利用计算流体动力学(CFD)方法,对某非能动三代反应堆蒸汽发生器U型管内流体的流动传热特性进行数值模拟分析。选取6组不同管长的U型管,对比分析U型管内单相流体的流动传热特性。基于数值仿真结果,得出6组U型管质量流量-进出口压降曲线,并?T分析了U型管长度和一次侧进口流体温度与二次侧壁面温度温差(?T)对流体回流的影响。研究结果表明,当?T一定时,随着进出口压降的降低,长管内更容易发生回流。当U型管长度一定时,?T越小越容易发生回流。 相似文献
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自然循环蒸汽发生器并联倒U型管流量分配计算 总被引:3,自引:3,他引:0
针对自然循环工况下蒸汽发生器部分倒U型管内存在倒流现象,通过对倒U型管内流动传热特性进行分析,获得了倒流发生的判断依据,从而编制了流量分配计算程序。采用该程序对某型蒸汽发生器并联倒U型管流量分配进行了计算,通过将结果与实验值进行对比分析,对程序可信度进行了验证,并采用该程序对蒸汽发生器并联倒U型管主要热工参数随进出口压降变化情况进行了计算分析。结果表明,倒流现象发生在短管内,倒流的发生使得蒸汽发生器一次侧净流量和单位时间输热呈阶梯下降,对反应堆安全产生较大的影响。 相似文献
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陈巧艳 《核工程研究与设计》2004,(49):18-21
在主回路冷段破口等效直径15.24cm的中破口失水事故分析,同时采用了不使用蒸汽冷凝回流模型、增大安注流量不使用蒸汽冷凝回流模型和使用蒸汽冷凝回流模型三种分析方法.分析结果表明:使用蒸汽冷凝回流模型时,回流的冷却剂可以有效地带走裸露燃料元件的热量,抑制燃料包壳温度升高.不使用蒸汽冷凝回流模型和增加安注流量时,裸露燃料元件的热量不能被带走,燃料包壳温度会升高. 相似文献
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对浮动式核电站中一类具有倾斜热管段的低压低高差自然循环系统的两相流动特性进行了实验研究,分析了加热功率对两相流动特性的影响。结果表明,不同功率条件下系统存在两相稳定冷凝和伴随蒸汽冷凝诱发水锤两相振荡2种流动模式,热管段内过冷水倒流和蒸汽与低温过冷水直接接触冷凝是导致2种流动模式的内在机制。此外,蒸汽冷凝诱发水锤的发生会产生较大压力脉冲,并导致过冷水倒流长度显著增加,进而加剧系统流动不稳定。进一步研究表明,加热段出口含气率可以作为流动不稳定判断依据。 相似文献
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由于阀门渗漏使核电厂安注系统冷水注入到充满热水的连接安注系统与主管道的支管中,而发生的热分层和温度振荡现象的研究对于确保核电厂的安全和可靠运行具有重要意义。运用计算流体力学软件CFX,采用k-ε湍流模型,以研究某核电厂安注系统支管中热分层现象的实验为对象,模拟了阀门渗漏冷水进入含有高温水的支管以后所发生的热分层现象,数值模拟的结果与实验测量结果吻合。在此基础上,通过改变阀门渗漏冷水的流量、支管的结构等参数,进一步研究支管中热分层现象与这些参数的内在关系,从而得出了影响热分层现象的主要原因及热分层现象发生的一些规律。 相似文献
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A potential cause of thermal fatigue failures in energy cooling systems is identified with cyclic stresses imposed on a piping system. These are generated due to temperature changes in regions where cold and hot flows are intensively mixed together. A typical situation for such mixing appears in turbulent flow through a T-junction, which is investigated here using Large-Eddy Simulations (LES). In general, LES is well capable in capturing the mixing phenomena and accompanied turbulent flow fluctuations in a T-junction. An assessment of the accuracy of LES predictions is made for the applied Vreman subgrid-scale model through a direct comparison with the available experimental results. In particular, an estimation of the minimal mesh-resolution requirements for LES is examined on the basis of the complementary RANS simulations. This estimation is based on the characteristics turbulent scales (e.g., Taylor micro-scale) that can be computed from LES or RANS simulations. 相似文献
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The mixing of coolant streams of different temperatures in pipe junctions leads to temperature fluctuations that may cause thermal fatigue in the pipe wall. Numerous T-junction experiments are known from literature, which were performed to study the nature of thermal loads in the pipe walls occurring during the mixing of hot and cold liquid. It is common to all known experiments that the experimental boundary conditions are set to reflect cases, in which the flow velocities in both main and side branches of the T-junctions are of the same order of magnitude. In the present experiments, carried out using wire-mesh sensors, it was observed that very low flow velocities in the side branch compared to the main pipe may lead to conditions potentially severe for thermal fatigue due to the low frequency of the temperature fluctuations occurring. The T-junction presented here consists of a perpendicular connection of two pipes of 50 mm inner diameter. The straight and the side branches are supplied with water of different electrical conductivities, to enable performing generic, isothermal tests on turbulent mixing with the idea to model the temperature fluctuations in thermal mixing processes. A pair of wire-mesh sensors, each with a grid of 16 × 16 measuring points, are used to record conductivity distributions in the downstream of the T-junction as well as directly at the junction in both branches. At very low flow rates in the side branch, a characteristic entrainment of liquid from the main branch into the side branch was found. Typically the entrainment flow in the side branch results in relatively high fluctuations at the low-frequency range. While the sensor in the main flow shows fluctuations with a power spectrum similar in character to mixing experiments with comparable flow velocities in both branches of the T-junction. The phenomenon of entrainment of water from the main branch into the side branch against the main flow direction vanishes at a certain critical velocity in the side branch. 相似文献
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The phenomenon of fluid/thermal mixing in the cold leg and downcomer of a Pressurized Water Reactor (PWR) has been a critical issue related to the concern of pressurized thermal shock. The question of imperfect mixing arises when the possibility of cold emergency core cooling water contacting the vessel wall during an overcooling transient could produce thermal stresses large enough to initiate a flaw in a radiation embrittled vessel wall. The temperature of the fluid in contact with the vessel wall is crucial to a determination of vessel integrity since temperature affects both the stresses and the material toughness of the vessel material. A simple mixing model is described which was developed as part of the EPRI pressurized thermal shock program for evaluation of reactor vessel integrity. 相似文献
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CARR应急堆芯冷却系统停堆冷却措施分析 总被引:1,自引:0,他引:1
停堆后的冷却问题是中国先进研究堆(CARR)重要的安全问题之一。CARR应急堆芯冷却系统是一套多功能、高度安全可靠的专设安全设施,它在反应堆正常运行时执行池水冷却功能;在正常停堆和事故停堆过程中执行应急堆芯冷却功能;还执行应急热阱选择、系统供电方式、回路阻力分析、阀门开关设置等方面的处理,使系统在两种功能的切换中不需要人为操作,依靠流量的自动匹配来满足正常运行和事故运行的要求。体现了CARR的安全性、先进性和经济性。本文以核安全法规和导则为前提,以满足系统功能为基础,首先介绍了CARR应急堆芯冷却系统的功能、主要参数和流程。根据CARR的实际情况,对应急堆芯冷却系统的停堆冷却措施和典型事故进行了分析,论证了该系统是如何在正常停堆和事故停堆状态下实现非能动堆芯冷却的。 相似文献
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Detailed simulation of the thermal stresses of the reactor pressure vessel (RPV) wall in case of pressurized thermal shock (PTS) requires the simulation of the thermal mixing of cold high-pressure safety injection (HPI) water injected to the cold leg and flowing further to the downcomer. The simulation of the complex mixing phenomena including, e.g., stratification in the cold leg and buoyancy driven plume in the downcomer is a great challenge for CFD methods and requires careful validation of the used modelling methods.The selected experiment of Fortum mixing test facility modelling the Loviisa VVER-440 NPP has been used for the validation of CFD methods for thermal mixing phenomena related to PTS. The experimental data includes local temperature values measured in the cold leg and downcomer. Conclusions have been made on the applicability of used CFD method to thermal mixing simulations in case with stratification in the cold leg and buoyant plume in the downcomer. 相似文献
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The paper deals with T-junction mixing experiments carried out with wire-mesh sensors. The mixing of coolant streams of different temperature in pipe junctions leads to temperature fluctuations that may cause thermal fatigue in the pipe wall. This is practical background for an increased interest in measuring and predicting the transient flow field and the turbulent mixing pattern downstream of a T-junction. Experiments were carried out at a perpendicular connection of two pipes of 51 mm inner diameter. The straight and the side branches were supplied by water of different electrical conductivity, which replaced the temperature in the thermal mixing process. A set of three wire-mesh sensors with a grid of 16 × 16 measuring points each was used to record conductivity distributions downstream of the T-junction. Besides the measurement of profiles of the time averaged mixing scalar over extended measuring domains, the high resolution in time and space of the mesh sensors allow a statistic characterization of the stochastic fluctuations of the mixing scalar in a wide range of frequencies. Information on the scale of turbulent mixing patterns is obtained by cross-correlating the signal fluctuations recorded at different locations within the measuring plane of a sensor. 相似文献
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Seyed Khalil Mousavian Francesco DAuria Mahmoud A. Salehi 《Nuclear Engineering and Design》2004,229(1):25-46
In all light water reactors (LWR), natural circulation is an important passive heat removal mechanism. In the present paper, the natural circulation phenomena are studied with reference to step-wise coolant inventory reduction and a small break loss-of-coolant-accident (SBLOCA) in the cold leg of VVER-1000. The natural circulation flow map (NCFM) approach is considered to evaluate the natural circulation performance of the VVER-1000 NPP also comparing VVER-1000 and PWR systems. Three different elevations between heat source (core) and heat sink (steam generators) zones have been considered in order to characterize the buoyancy force in a VVER-1000. The influence of power and the cold legs loop seal upon the natural circulation performance is also evaluated. In the second part, a series of SBLOCA simulations with break area ranging from 0.5 to 11.7% of the cold leg cross sectional area are performed starting with the VVER-1000 system in nominal conditions. The effect of Emergency Core Cooling System (ECCS) including passive and active parts of ECCS are evaluated. The simulations were performed by the help of the system code RELAP5. Within the framework of the qualification of the adopted computational tools, the results are compared with experimental data from Kozloduy NPP unit 6 test and PSB-VVER integral test facility available from the literature. Namely, the qualification of the adopted nodalisation in steady state conditions is achieved by using experimental data. The accuracy of selected results have been estimated in quantitative terms by applying the fast Fourier transform based method (FFTBM). Finally, the relevance and the potential for the occurrence of the reflux condensation mode, i.e., one of the Natural Circulation regimes, for cooling of reactor core in VVER-1000 are discussed. 相似文献