共查询到16条相似文献,搜索用时 203 毫秒
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为了研究压水堆因安注冷水直接注入反应堆压力容器下降环腔而导致的承压热冲击(PTS)热工水力问题,基于1∶10比例模型,应用计算流体力学商用软件FLUENT5.4进行了紊流流动换热的数值模拟分析,同时进行了常压传热实验研究。针对下降环腔折算流速0.5m/s,安注流速10m/s的典型工况,研究了压力容器下降环腔的壁面换热特性。通过分析下降环腔内的流动及混合特性,从流动机理上解释了压力容器内壁上准重接触点附近换热强烈的现象,并指出壁面换热强弱与近壁流体紊流脉动动能密切相关,为热冲击分析提供参考。 相似文献
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失水事故工况 (LOCA)下反应堆下降环腔内的流动和传热研究 ,对反应堆压力容器 (RPV)的安全具有重要的意义。通过对一种直接安注的反应堆压力容器内流动和传热的研究 ,将流动分为横穿射流和冲击射流 ,比较了在两种射流下下降环腔内流动和传热的特点 ,分析了流速比和对流换热系数及温度的关系 ,当流速比在 1~ 1 0时 ,流动属于横穿射流 ,对流换热主要由环腔流速决定 ;流速比大于 1 0后 ,属于冲击射流 ,环腔内对流换热主要决定于安注流速 ,此时局部对流换热能力随安注流速的增加而增加 相似文献
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带热套管的T型接管内流动换热的数值模拟和实验研究 总被引:1,自引:0,他引:1
为了分析核反应堆冷却剂系统中带热套管T型接管内由于注入非等横向射流导致的构件热冲击状况,本文应用计算流体力学商用软件FLUENT5.3进行了紊流流动换热的数值模拟,分析了主管及接管与热套间环腔内的流动换热特性,针对套管上开有通流小孔,并采用凸台支撑的热套管结构形式,模拟了射流与主流流速比为0.05及0.5两种典型工程,传热实验,研究了主管及接管内壁近壁区域的传热特性,并讨论了热套管尺寸变化对接管热冲击的影响,结果表明,数值模拟与实验数据吻合良好,热套管对构件的热保护程度与热套管结构形式及流速比密切相关,适当减小流速比有利于改善构件热应力状况。 相似文献
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我国非能动系列压水堆将应急冷却系统冷却水的注入管道直接连接于压力容器上,与传统的冷管段安注不同,这种安注方式被称之为反应堆压力容器直接安注。本文以安注条件下的反应堆压力容器为研究对象,采用物理实验与数值分析结合的方法,对安注流体在压力容器表面形成的热分布形态进行研究。研究发现,不同于传统的主管道冷段斜接管安注方式,直接安注条件下安注流体在下降环腔中的分布形态接近于等腰三角形。以实验结果为基础,结合数值计算验证,发现了压力容器热分布角与流速比成正比关系,并进一步提出了安注流体分布计算模型,从而为反应堆安全设计提供参考。 相似文献
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为探究反应堆压力容器下降段在喷放末期冷段安注过程中的水-蒸汽逆流特性,建立下降段逆向流动限制(CCFL)模型,开展了基于压力容器模化本体的下降段CCFL实验研究以及建模分析。通过实验研究获得了不同入口安注水流量、安注水过冷度、堆芯蒸汽流量等条件下的下降段环腔内的安注特性数据,并基于实验数据进行了CCFL建模分析。结果表明,开始发生CCFL的蒸汽无量纲流速与入口安注水无量纲流速呈现正相关,基于无量纲流速建立的模型斜率与入口安注水无量纲流速呈现高度指数关联。本文建立了适用于从不发生CCFL至不完全CCFL,再到完全CCFL的下降段水-蒸汽气液逆流全过程预测模型。 相似文献
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为探究工质在核热推进反应堆冷却剂通道内的热工水力行为,基于数值计算方法,开展了圆管内高温、高流速氢气流动换热特性研究。通过与实验数据对比发现,采用压力基耦合算法、SST k-ω湍流模型以及物性模型进行高温、高流速氢气流动换热特性数值模拟是合理可行的,计算值与实验值符合较好,计算模型选择正确。在分析基础工况流场与温度场的基础上,还研究了热工参数对氢气管内流动换热特性的影响,结果表明,随质量流量的增大换热效果增强,随热流密度的增大换热效果变差。研究方法与结果可为高温、高热流密度环境下气体工质流动换热特性研究、核热推进反应堆的热工设计与仿真模拟提供参考。 相似文献
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The object of this work is to investigate fluid mixing phenomena as they related to pressurized thermal shock (PTS) in a pressurized water reactor vessel downcomer during transient cooldown with direct vessel injection (DVI), using test models. The test model designs were based on ABB Combustion Engineering (CE) System 80+ reactor geometry. A cold-leg, small-break loss-of-coolant accident (LOCA) and a main steam line break were selected as the potential PTS events for the ABB-CE System 80+. This work consists of two parts. The first part provides the visualization tests of the fluid mixing between DVI fluids and existing coolant in the downcomer region, and the second part presents the results of thermal mixing tests with DVI in the other test model. Flow visualization tests with DVI have clarified the physical interaction between DVI fluid and primary coolant during transient cooldown. A significant temperature drop was observed in the downcomer during the tests of a small-break LOCA. The measured transient temperature profiles compare well with the predictions from the REMIX code for a small-break LOCA, and with the calculations from the COMMIX-1B code for a stream line break event. 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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The paper describes analyses performed with the Reactor Excursion and Leak Analysis Package 5 (RELAP5) computer code to investigate pressurized thermal shock transients for the H.B. Robinson pressurized water reactor. The computer models and their application to 180 transients are described. Reactor vessel downcomer temperature and pressure histories for five transient groups are presented. 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 phenomenon of the pressurized thermal shock on the reactor pressure vessel is expected to occur in the case of such an accident as the small loss of the coolant accident in the PWR nuclear plant. In order to study the structural integrity of the reactor pressure vessel under the pressurized thermal shock, the cleavage thermal shock fracture experiment was conducted here using an initially corner-cracked nozzle type specimen made of the pressure vessel steel A508 class 3. The fracture mechanics analysis was performed to asses bthe crack behaviors in the experiment using the time dependent stress intensity factor deduced from the three-dimensional J integral with the thermal effect. 相似文献