共查询到18条相似文献,搜索用时 209 毫秒
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热工水力数值模拟是反应堆系统设计和安全分析的重要内容,以RELAP5为代表的系统程序可对瞬态或事故工况进行快速分析,同时以FLUENT为代表的计算流体动力学(CFD)程序对堆芯局部三维现象的分析也越来越重要。为综合利用两者的优点,以RELAP5/FLUENT为基础,利用对RELAP5程序源代码的二次开发和FLUENT的用户自定义函数(UDF)进行编程,开发了RELAP5/FLUENT耦合程序。利用flibe熔盐在水平圆管流动问题验证了程序耦合的正确性;针对2 MW熔盐堆进行了稳态模拟,耦合程序能详细分析熔盐堆的热工水力行为;模拟了2 MW熔盐堆功率突变的瞬态热工水力行为,相对于单独的RELAP5,耦合程序能更好地揭示熔盐堆系统和堆芯的三维物理现象。该耦合程序可用于解决熔盐堆热工水力分析中存在的显著三维混合现象的问题。 相似文献
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《核科学与工程》2018,(5)
本文为了更加真实准确的模拟非能动核电机组复杂的热工水力工况,提高事故分析计算精度,开展了非能动压水堆热工水力多尺度耦合计算分析研究。首先应用热工水力系统分析程序RELAP5对AP1000机组进行系统建模,并开展冷却剂强迫流动完全丧失事故(全失流事故)的分析计算,得到堆芯相关热工水力参数。然后将RELAP5程序的计算结果作为边界条件,分别利用子通道程序COBRA-Ⅳ、计算流体力学程序FLUENT以及基于两个程序的耦合程序对AP1000堆芯组件进行建模,并分别开展全失流事故过程中堆芯热工水力分析计算。最终通过三个程序计算结果的对比,表明应用耦合程序开展堆芯热工水力分析的方法可行,建立的堆芯组件模型合理,计算结果更加接近真实情况,有效减少了单一程序计算的过度保守性。 相似文献
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在系统热工水力程序RELAP5/mod3.2的基础上,采用显式方法建立了堆芯三维时空中子动力学与一维热工水力计算的耦合模型,接入基于非线性迭代半解析节块法的三维瞬态物理分析模型(NLSANMT)后,形成了一个具有堆芯三维瞬态物理特性分析能力的系统计算程序NLSANMT/RELAP5(mod3.2).通过核动力反应堆温度反馈系数、堆芯功率分布参数的校算及单束控制棒失控抽出事故的模拟分析,验证了接口的正确性.验证结果表明,与RELAP5/mod3.2相比,所开发的NLSANMT/RELAP5(mod3.2)程序具有更强的堆芯物理瞬态分析能力. 相似文献
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基于二次开发得到的铅冷快堆一维系统程序RELAP5_LEAD和三维计算流体力学程序FLUENT,利用动态链接库技术和FLUENT用户自定义函数,开发了多尺度耦合分析程序RELAP5/FLUENT。在单相范围内,分别利用耦合程序RELAP5/FLUENT开展简单铅冷串联管道的瞬态流动和传热模拟、简单铅冷闭式回路的瞬态流动模拟,并与RELAP5_LEAD计算结果开展Code-to-Code对比分析。研究结果表明,RELAP5/FLUENT计算结果与RELAP5_LEAD模拟结果吻合良好,耦合程序的开发取得了初步成功,可用于分析铅冷快堆堆内的复杂三维热工水力现象。 相似文献
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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。 相似文献
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为研究西安脉冲堆(XAPR)在意外引入反应性且停堆系统失效事故下的瞬态安全特性,本文基于XAPR的结构和运行特点,建立了适用于XAPR的瞬态热工水力分析模型,并开发了用于XAPR安全特性分析的瞬态热工水力程序TSAC-XAPR。利用TSAC-XAPR程序对反应性引入事故进行模拟计算,结果表明:当XAPR在额定功率范围内运行时,发生反应性引入事故后,堆芯能依靠自身的固有反馈机制使脉冲堆重新达到稳定运行状态;当运行功率过高尤其是超过临界值时,反应性引入事故将导致脉冲堆关键热工水力参数发生振荡,无法再次达到稳态。此外,不同反应性引入方式将影响堆芯参数在反应性引入过程中的变化趋势,但并不影响其最终稳态值。 相似文献
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为研究铅铋快堆瞬态热工水力特性,对RELAP5程序进行二次开发,添加铅铋合金(LBE)物性模型和液态金属流动换热模型,并与NACIE-UP和CIRCE-ICE台架的实验结果进行对比。计算结果表明:NACIE-UP台架稳态流量和温度相对误差在2%以内,瞬态相对误差不超过5%,与其他系统程序CATHARE、ATHLET、RELAP5-3D、RELAP5/MOD3.3(modified)相比,本文程序的相对偏差不超过10%;CIRCE-ICE台架稳态流量和温度相对误差在2%以内,瞬态相对误差不超过10%。本文程序满足反应堆系统热工水力分析程序精度要求,可作为铅铋快堆安全分析的有效工具。 相似文献
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Pressurized water vessel-type reactor (VVER) safety has become a very important issue, in particular for countries in Central and Eastern Europe. For thermal-hydraulic analyses the western codes like RELAP5, CATHARE and ATHLET were used.The purpose of the study was to quantitatively assess the RELAP5 capability to predict the main circulation pump (MCP) trip at nearly full power transient in Mochovce VVER 440/213 nuclear power plant (NPP). The transient parameters were recorded during the start up test program implementation. For accuracy quantification the improved fast Fourier transform based method (FFTBM) was used. The RELAP5/MOD3.2.2 computer code was used for calculation. The results showed very good agreement between calculated and plant measured data. The results also confirmed some previous studies that the simpler is the transient the higher code accuracy is generally achieved. 相似文献
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TIAN Xiaoyan CHEN Sen YANG Ning ZHU Lei LI Huaqi MA Tengyue HU Pan KANG Xiaoya 《原子能科学技术》1959,54(11):2089-2097
In order to study the transient safety characteristics of Xi’an Pulsed Reactor (XAPR) when unexpected reactivity insertion accident happened and shutdown system failed, the main mathematical models were established based on the specific core structure and operation conditions of XAPR. Meanwhile, a transient thermal-hydraulic code called TSAC-XAPR was developed to analyze the safety characteristics of XAPR. The TSAC-XAPR code was then used to simulate the reactivity insertion accident of XAPR. The calculation results indicate that when XAPR operating under rated power, reactor can reach a new steady state for reactivity insertion accident, depending on its inherent feedback mechanism. When XAPR operating under high power, especially above the critical power, key thermal-hydraulic parameters of reactor will tend to oscillate and can’t reach a steady state again for reactivity insertion accident. Besides, it is also found that different reactivity insertion modes will only affect the variation trend during the phase of reactivity insertion instead of the final value at steady state. 相似文献
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基于RELAP5 MOD3.2的钠冷快堆热工水力系统分析程序开发及验证 总被引:1,自引:1,他引:0
对大型核反应堆热工水力分析程序RELAP5 MOD3.2进行了改造,使之适用于钠冷快堆系统安全分析。在不影响原程序功能的基础上添加了气液两相钠物性和液态金属对流换热模型,并改造了相应的初始化模块和计算模块。改造后的程序可正确模拟钠的流体力学特性和热物性,搭建钠冷快堆热工水力流体网络进行分析计算。对EBR-Ⅱ试验堆基准题进行了稳态模拟和失流事故分析,其中稳态计算主要参数与实验值相对偏差小于1%,瞬态计算相对偏差小于10%,各参数变化趋势与实验值相符良好,初步验证了改造程序的可靠性。 相似文献
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《Progress in Nuclear Energy》2012,54(8):1095-1104
Nowadays, new concepts of nuclear reactors have been projected to work with mechanisms of natural circulation (NC). However, NC systems are very susceptible to several kinds of instabilities being necessary careful studies about such systems. In this work, a theoretical investigation about BWR stability during a transient of recirculation pump trip bringing the reactor to operate at NC conditions is presented. The simulations were performed using the RELAP5/MOD3.3 thermal-hydraulic code and the PARCS/2.4 3D neutron-kinetic code in a coupled way to predict the transient results. The power time evolution and the related thermal-hydraulic parameters were investigated during the transient to analyze the behavior of the reactor for this special operation condition of NC. 相似文献
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Humberto V. Soares Antonella L. Costa Claubia Pereira Maria Auxiliadora F. Veloso Patrícia A.L. Reis 《Progress in Nuclear Energy》2011,53(8):1095-1104
Nowadays, new concepts of nuclear reactors have been projected to work with mechanisms of natural circulation (NC). However, NC systems are very susceptible to several kinds of instabilities being necessary careful studies about such systems. In this work, a theoretical investigation about BWR stability during a transient of recirculation pump trip bringing the reactor to operate at NC conditions is presented. The simulations were performed using the RELAP5/MOD3.3 thermal-hydraulic code and the PARCS/2.4 3D neutron-kinetic code in a coupled way to predict the transient results. The power time evolution and the related thermal-hydraulic parameters were investigated during the transient to analyze the behavior of the reactor for this special operation condition of NC. 相似文献