CEFR钠空泡反应性效应试验测量与计算分析

本文介绍了中国实验快堆物理启动试验中钠空泡反应性效应测量试验的试验程序及测量结果评估,测量结果显示中国实验快堆典型位置钠空泡反应性价值皆为数值较大的负反应性,结果符合试验验证要求,验证了组件瞬间堵流事故专设监测系统的信号基础。并对试验进行了计算分析,试验前的分析为试验提供支持,试验验证了计算分析程序系统。     

多孔介质方法在池式快堆系统分析软件SAC-CFR三维钠池计算模型中的应用

为准确分析池式快堆热钠池内的热工水力学特性,在已开发出的用于池式快堆系统分析的钠池三维计算模型的基础上,应用多孔介质方法建立钠池内中间热交换器、主泵、事故热交换器及屏蔽柱模型,完成了含有多孔介质模型和复杂几何边界的钠池三维计算模型开发。将该模型嵌入池式快堆系统分析软件SAC-CFR后,分析了中国实验快堆在稳态运行和紧急停堆工况下钠池内的流场分布,初步证明了所采用的多孔介质模型的合理性,为下一步非能动余热排出系统模型的开发做准备。     

粘钠设备水蒸气氮气清洗工艺研究

快堆电站一、二回路的设备因为粘有冷却剂—钠,在维修和退役前必须进行除钠处理。水蒸气氮气清洗是快堆电站设备清洗系统所采用的清洗除钠工艺。本文首先对该工艺的清洗原理进行了分析,然后对工艺研究的试验装置进行了论述,最后通过两种特征的粘钠设备的清洗,深入分析清洗过程中所出现的各种现象,为后续的快堆电站粘钠设备清洗技术的研究提供了保证,同时获得了可直接用于中国实验快堆(CEFR)设备清洗系统的运行经验。     

粘钠设备水蒸气氮气清洗工艺研究

快堆电站一、二回路的设备因为粘有冷却剂——钠,在维修和退役前必须进行除钠处理。水蒸气氮气清洗是快堆电站设备清洗系统所采用的清洗除钠工艺。本文首先对该工艺的清洗原理进行了分析,然后对工艺研究的试验装置进行了论述,最后通过两种特征的粘钠设备的清洗,深入分析清洗过程中所出现的各种现象,为后续快堆电站粘钠设备清洗技术的研究提供了保证,同时获得了可直接用于中国实验快堆(CEFR)设备清洗系统的运行经验。     

中国实验快堆蒸汽发生器首次进水试验分析研究

中国实验快堆(CEFR)作为中国第一座钠冷快中子反应堆,蒸汽发生器作为分隔二回路和三回路器的重要设备,其运行的稳定性、可用性对于中国实验快堆的稳定运行具有重要意义。中国实验快堆蒸汽发生器首次进水试验验证了蒸汽发生器具备了在低功率下运行的完整性、可用性、稳定性和良好的传热性能。根据记录数据就钠侧和水侧的对流换热进行了热平衡校核计算,计算结果表明了本次试验钠侧和水侧换热量平衡,计算结果表明在低流量、低钠温的运行工况下水侧为主要换热热阻,该热阻值可以由格尼林斯基(Gnielinski)公式确定。     

钠冷快堆中安全棒落棒过程的水力计算

安全棒及其驱动机构是钠冷快堆中的紧急停堆系统,其落棒时间一直是钠冷快堆安全设计关注的重点之一。本文利用一维水力管网方法对中国实验快堆中的安全棒系统落棒过程开展了详细的建模和分析,得到了安全棒在落棒过程中的位移、速度、加速度与时间的关系,并与中国实验快堆调试试验结果进行了对比。同时,基于该方法开展了中国实验快堆安全棒落棒时间与冷却剂流量及温度的敏感性分析及研究。     

中国实验快堆泵支承冷却系统温度场分析

中国实验快堆一回路泵支承套筒是承重设备,位于高温的热钠池中.为了限制套筒和套简内部冷钠腔室的钠温度,维持主泵正常工作温度,设置钠泵支承冷却通道.利用计算流体动力学技术(CFD),对泵支承冷却系统进行三维模拟,通过对泵支承冷却系统冷却流道和支承结构的数值传热分析,得到了该系统的温度场分布情况,验证了泵支承冷却系统的冷却能力.     

快堆热钠池的三维热工水力分析

采用多孔介质的方法，对中国实验快堆（ＣＥＦＲ）热钠池的复杂流场进行了数值模拟，计算结果与设计值符合得很好，并给出了流场分布，为中国实验快堆的进一步设计和安全性能评价提供了重要的参数。     

池式快堆三维热工流体力学设计研究

随着计算机软硬件技术的发展,三维数值分析技术已经成为池式快堆堆芯和钠池热工设计和计算分析的重要组成部分,并在其中发挥着不可替代的作用.通过对池式快堆几个典型热工现象的分析,展示了我国第一座池式快堆(中国实验快堆)热工设计和安全分析中所拥有的设计手段和工具,总结了三维数值分析技术在快堆工程中的应用,并指出了其对今后快堆热工设计的重要意义.     

基于附加源项法的钠冷快堆冷热池三维分析

选择计算流体动力学(CFD)为模拟手段,建立快堆一回路钠池的三维闭式一体化CFD模型,对一回路中主要部件进行模拟,其中中间热交换器、独立热交换器、堆芯、主泵采用附加源项法进行模拟,得到中国实验快堆(CEFR)额定功率稳态运行时整个流场的三维速度场与温度场。计算值同CEFR设计值进行比较,结果符合预期,证明了模型的合理性。计算结果表明,钠池较明显地分为温度较低的冷钠池和温度较高的热钠池2个部分,热钠池温差较大,冷热流体搅混现象明显;同时冷钠池、热钠池不同高度的平均温度都很接近,说明分隔冷热钠池的热屏蔽效果较好。     

某核电厂主蒸汽管道应力分析

本文研究了某核电厂中主蒸汽系统管道的计算和评定等典型内容。此系统管道运行中承受的载荷工况多样,管道应力状态复杂。为了保证系统管道能够正常运行,在设计上需保证该系统管道的应力能够满足相关规范要求。分析采用管道力学分析软件PIPESTRESS进行,计算模型包括主回路、主蒸汽系统及相关的管道和阀门,分析包含静力和动力计算等。对计算结果依据美国机械工程师学会的ASME及相关规范进行了应力评定,并包含了LBB评定,保证了回路运行的安全。     

A study on the phenomenon of rate of loading in motor operated gate valves

This study analyzed the rate of loading (ROL) phenomenon, which is generated during the operation of a motor operated valve (MOV) under fluid pressure conditions. ROL is one of the most important parameters for an MOV performance evaluation. This paper includes the analysis results for the characteristics of ROL and the effect of fluid pressure on the ROL. Dynamic and static test were performed to analyze the ROL effect for flexible wedge gate valve. The result of this analysis confirmed that the ROL is generated under fluid pressure condition and that the ROL value under high differential pressure condition appeared to be higher than under low differential pressure condition. According to the test results of multiple valves, the ROL appeared to become higher, as the differential pressure increased, and under the high differential pressure condition, it accounted for approximately 17.6% of the thrust loss. In addition, the ROL effect was negligible in valves with a low differential pressure (below 1100 kPa).     

百万千瓦级压水堆严重事故卸压阀高温瞬态分析

由于核电厂严重事故的恶劣工况,在卸压过程中严重事故卸压阀门可能会经历阀门无法承受的高温瞬态而导致不可用。本文在可能导致高压熔堆的事故序列中筛选出具有一定的包络性并包含各种典型严重事故现象的典型严重事故序列。针对该事故序列考虑严重事故管理中的开阀时间范围开展了高温瞬态计算,并针对重要的影响因素阀门开启时刻的稳压器水位开展分析。最终确定了百万千瓦级核电厂具备典型性及一定包络性的严重事故卸压阀工作条件,并得到了阀门开启前后阀门可能经历的最高流体温度及流体温度变化曲线,为严重事故卸压阀门的设备鉴定及功能应用提供了重要基础。      

核电阀门国产化研究

介绍了我国核电站阀门的基本情况、工况和技术要求以及我国核电阀门采用的技术标准.说明了当前核电站阀门的发展趋势和我国核电阀门国产化现状.分析了当前我国核电阀门国产化率偏低的原因,指明了当前我国核电阀门国产化存在的不足之处,编者就实现核电阀门国产化的基本目标提出了的自己的建议和措施.     

Thermal-Hydraulic System Study of the Helium Cooled Pebble Bed （HCPB） Test Blanket Module （TBM） for ITER Using System Code RELAP5

The HCPB concept has been a European DEMO reference concept for nearly one decade. Detailed thermal-hydraulic study on the control behavior of the whole system is one of the important parts of this development. The thermal-hydraulic effect of the TBM-combined cooling circuit during a cyclic operation in ITER has been studied using the system code RELAP5. The RELAP5 is based on an one-dimensional, transient two-fluid model for the flow of a two-phase steam-water mixture that can contain noncondensable components like Helium. The RELAP5-models are modified to take the cyclic operation of the circulator, heat exchanger, bypass, valves etc in to account. A sequence of operational phases is investigated, starting from the cold state through the heating phase that brings the system to a stand-by condition, followed by typical power cycles applied in ITER. The results show that the implemented control mechanisms keep the inlet temperature to the TBM and the total mass flow rate at the required values through all phases.     

Analysis of flow induced valve operation and pressure wave propagation for single and two-phase flow conditions

The flow induced valve operation is calculated for single and two-phase flow conditions by the fluiddynamic computer code DYVRO and results are compared to experimental data. The analysis show that the operational behaviour of the valves is not only dependent on the condition of the induced flow, but also the pipe flow can cause a feedback as a result of the induced pressure waves. For the calculation of pressure wave propagation in pipes of which the operation of flow induced valves has a considerable influence it is therefore necessary to have a coupled analysis of the pressure wave propagation and the operational behaviour of the valves.The analyses of the fast transient transfer from steam to two-phase flow show a good agreement with experimental data. Hence even these very high loads on pipes resulting from such fluiddynamic transients can be calculated realistically.     

Sodium Fire Accident Scenario Simulation Technology Study Based on FDS Code

Sodium fire is a typical accident of sodium-cooled fast reactor. Simulation of sodium fire accident scenario by software is a powerful tool for risk assessment of sodium fire accident. In this paper, the conventional fire three-dimensional computational fluid dynamics software FDS was used as a platform to add a sodium fire combustion model, including combustion heat model, combustion rate model, spray liquid sodium particle size distribution model, etc., and complete the development of sodium fire scenario modeling analysis program. And through the comparison with SPHINCS sodium fire test and calculation results, the feasibility of the method and development plan was verified. The research results of this paper can provide the research basis and experience reference for the development of the subsequent sodium fire simulation program.     

基于FDS程序的钠火事故情景模拟技术研究

钠火是钠冷快堆的典型事故，钠火事故情景计算机模拟仿真是对钠火事故风险评价的有力工具。本文以常规火灾三维计算流体力学软件FDS为平台，增加钠火燃烧模型，包括燃烧热模型、燃烧速率模型、喷射液钠粒径分布模型等，完成了钠火情景模型的开发，并通过与SPHINCS钠火试验和计算结果的温度分布与氧气含量对比，验证了模拟技术和模型开发方案的可行性。本文的研究成果能为后续钠火仿真模拟程序的开发提供研究基础和经验参考。     

钠冷快堆钠火概率安全评价方法研究

钠火事故是钠冷快堆的典型和特有事故,且很可能是反应堆总风险的主要贡献因素之一。本文在介绍钠火事故特点的基础上,研究使用概率安全分析评价钠冷快堆钠火风险的方法。以中国实验快堆反应堆大厅钠火事故为实例,计算得到反应堆大厅钠火导致的堆芯损坏频率为1.19×10^-8/(堆•年)。在此基础上进一步讨论目前钠火概率安全评价中尚需研究的关键问题。     

Dynamic Testing and Seismic Qualification Practice: Clarence W. de Silva (Lexington Books,D.C. Heath and Company,Lexington, Massachusetts,USA)

The closing valves are the main part of time in stand-by state. It is characteristic that the reliability of the valves decreases during stand-by time. Therefore the valves are tested periodically in order to check their condition. On the other hand, the testing may cause failures and thus decrease the availability of valves. Through a proper choice of test internal these competing factors can be balanced against each other. Optimum tests interval of closing valves is considered with the aid of simple unavailability models and using reliability data collected from Finnish BWR plants.