10MW高温气冷实验堆吸收球停堆系统的设计

介绍了吸收球停堆系统的设计原则，分析了不同参数对系统设计的影响，并对吸收球停堆系统的最大可信事故进行了分析。分析表明，本吸收球停堆系统的设计能实现在任何工况下的启动和运行，不会发生失效。     

10 MW高温气冷实验堆硼吸收球停堆系统气力输送模拟试验研究

吸收球停堆系统是10MW高温气冷实验堆(HTR-10)的第二停堆系统,于紧急事故停堆之后、重新开堆之前投入运行,利用负压输送过程将在紧急停堆时进入反应堆堆芯落球孔道内的中子吸收球输送到位于堆顶的贮球罐内,实现正常开堆或反应堆再临界。运用气力输送的密相输送理论,对回路各部件和各管段的气固两相流阻力进行计算,并在1：1模拟试验台架上,以空气和氦气为载体,真实硼吸收球为物料,进行了气力输送试验研究。试验数据与理论分析相符合,吸收球第二停堆系统的气力输送功能满足HTR-10工程的技术要求。     

气力输送在10MW高温气冷实验堆吸收球停堆系统中的应用

吸收球停堆系统是10MW高温气冷实验堆（HTR－10）的第二停堆系统,本系统利用负压输送过程,将在紧急停堆时进入堆芯反射层孔道内的B4C小球输送到位于堆顶的贮球罐内,实现正常开堆或反应堆再临界。气力输送技术应用于反应堆工程是一种新的尝试与探索,由此带来了一一些新的课题有待进一步的理论和实验研究。本文运用密相输送理论计算了气力输送过程中氦气与B4C球的气固两相流阻力,并对该系统的最大可信故障进行了理论分析,为该系统的参数设计及设备选型提供了理论依据。     

HTR-10第二停堆系统的试验与调试

吸收球停堆装置是10MW高温气冷实验堆的第二停堆系统，控制棒失效时，碳化中子吸收球落入堆芯反层的吸收球孔道内，实现紧急停堆；反应堆再次临界前，利用气体输送装置将吸收球送回位于堆顶的贮球罐内，在实验室和高 温堆上先后进行了7套吸收球装置的热态试验和输送功能试验，试验数据表明，吸收球系统7套装置的落球和回球动作正常，所用的时间在要求的范围内；球位状态指示正常；气体回路流动正常，风机的流量，压升正常，12个阀门的开，闭功能正常。     

10MW高温气冷堆热气导管高温性能试验

水平布置的同轴双层热气导管在１０ＭＷ在高温气冷实验反应堆中是连接堆芯和蒸汽发生器的重要部件， 外分别流过高温和低温氦气，在氦气工程试验回路上进行了热气导管热工性能试验，使用氦气介质，在３．０ＭＰａ，９５０℃温度下连续运行时间超过９８ｈ，ｄ３．０ＭＰａ７００℃以上温度条件下的热运行时间超过３５０ｈ，还在０．１－３．４ＭＰａ压力范围内进行了２０闪压力循环；在１００－９５０℃范围内，进行了１８次温度循     

10MW高温气冷实验堆的堆体结构特点

模块式高温气冷堆是当今世界上公认的先进反应堆堆型之一。固有安全性是它的最突出的优点。本文对10MW高温气冷堆的堆体布置进行了详细描述,并对10MW高温气冷堆的结构设计特点进行了分析。根据10MW高温气冷堆的特点,本文对该堆的固有安全性、制造工艺等方面的优点进行了论述。     

10MW高温气冷实验堆初装堆方案设计初步设计

设计了１０ＭＷ高温气冷实验初装堆的两个方案，采用高温气冷堆物理设计程序包ＶＳＯＰ其进行分析计算，结果表明；两方案均能实现比较平稳地向平衡态过渡。就过渡过程中的单球最大功率、最大燃耗等参数而言，方案２优于方案１。     

吸收球停堆系统驱动装置螺栓抗震初步分析

吸收球停堆系统在高温气冷堆中起到相当重要的反应性控制和调节作用。而驱动装置是吸收球停堆系统中控制吸收球下落的关键运动部件。高约5m、呈细长结构的吸收球停堆系统驱动装置通过贮球罐底座与金属堆内构件的上支承板安装面相连。吸收球停堆系统贮球罐和驱动机构均为抗震Ⅰ级设备,故驱动装置连接螺栓的抗震校核计算是非常重要的。在本文中,通过将复杂的驱动装置简化为3段变截面结构,分析结构的超静定问题,对驱动装置内贮球罐底部与顶部的螺栓进行了校核计算。计算结果表明:贮球罐底部与顶部螺栓均在抗拉强度的安全范围内,同时给出了驱动机构薄弱处的支承力。     

10MW高温气冷实验堆电力系统

10MW高温气冷堆电力系统是为保障反应堆完成其安全目标和运行目标而设置的重要系统,本文介绍了10MW高温气冷堆电力系统的组成结构和配置,描述了电力系统在不同工况下的运行方式。     

高温气冷堆备用停堆系统电磁铁驱动机构初步研究

介绍了高温气冷堆备用停堆系统及其可选的驱动机构,分析了电磁铁驱动机构的工作原理和设计要求,并通过经验公式方法和专业软件来计算电磁铁线圈参数,利用模型实验检验两种计算方法的可靠性.通过软件数值计算方法得出了满足电磁铁吸力要求的线圈安匝数,并获得了启动过程的吸力特性曲线.     

Design of the material performance test apparatus for high temperature gas-cooled reactor

Most materials can be easily corroded or ineffective in carbonaceous atmospheres at high temperatures in the reactor core of the high temperature gas-cooled reactor （HTGR）. To solve the problem, a material performance test apparatus was built to provide reliable materials and technical support for relevant experiments of the HTGR. The apparatus uses a center high-purity graphite heater and surrounding thermal insulating layers made of carbon fiber felt to form a strong carbon reducing atmosphere inside the apparatus. Specially designed tungsten rhenium thermocouples which can endure high temperatures in carbonaceous atmospheres are used to control the temperature field. A typical experimental process was analyzed in the paper, which lasted 76 hours including seven stages. Experimental results showed the test apparatus could completely simulate the carbon reduction atmosphere and high temperature environment the same as that confronted in the real reactor and the performance of screened materials had been successfully tested and verified. Test temperature in the apparatus could be elevated up to 1600℃, which covered the whole temperature range of the normal operation and accident condition of HTGR and could fully meet the test reauirements of materials used in the reactor.     

Analysis on thermophoretic deposit of fine particle on water wall of 10 MW high temperature gas-cooled reactor

The water wall is an important part of the passive natural circulation residual heat removal system in a high temperature gas-cooled reactor. The maximum temperatures of the pressure shell and the water wall are calculated using annular vertical closed cavity model. Fine particles can deposit on the water wall due to the thei‘mophoresis effect. This deposit can affect heat transfer. The thermophoretic deposit efficiency is calculated by using Batch and Shen‘s formula fitted for both laminar flow and turbulent flow. The calculated results indicate that natural convection is turbulent in the closed cavity. The transient thermophoretic deposit efficiency rises with the increase of the pressure shell‘s temperature. Its maximum value is 14%.     

Thermal mixing characteristics of helium gas in high-temperature gas-cooled reactor, (I) thermal mixing behavior of helium gas in HTTR

The future high-temperature gas-cooled reactor (HTGR) is now designed in Japan Atomic Energy Agency. The reactor has many merging points of helium gas with different temperatures. It is needed to clear the thermal mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the high-temperature engineering test reactor (HTTR) due to lack of mixing of helium gas in the primary cooling system. Now, the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal–hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the thermal mixing behavior of helium gas. As a result, it was confirmed that the thermal mixing behavior is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR.     

Conceptual design and analysis of a passive residual heat removal system for a 10 MW molten salt reactor experiment

A conceptual design of a passive residual heat removal system was developed for a 10 MW molten salt reactor experiment (MSRE) designed by Oak Ridge National Laboratory (ORNL). The principle, main components and design parameters of the system were presented, and thermal-hydraulic behaviors, such as natural circulation and heat removal ability, were numerically analyzed in the code of C++, especially for the bayonet cooling thimbles. The results show that the system can effectively remove decay heat in the molten salt in an MSRE and has a heat removal rate that approximates to the decay heat generation rate, thus causing the temperature of the molten salt to decrease steadily. The width of the gas gap in the bayonet cooling thimbles has little effect on either the heat exchange or the natural circulation inside the thimbles, while the width of the steam riser, in spite of its slight effect on the heat transfer of the system, greatly influences the natural circulation. With the width of the steam riser increase from 3.6 to 5.1 mm, the mass flow rate increases from 1.9 kg/s to 4.79 kg/s. Finally, three operational schemes were proposed for the passive residual heat removal system, among which that of reducing the bayonet cooling thimbles by three-quarters had the best comprehensive performance.     

A model for estimation of reactor spurious shutdown rate considering maintenance human errors in reactor protection system of nuclear power plants

Modeling of spurious activations in safety instrumented systems has been studied for over a decade. The spurious activation of a plant protection system in nuclear power plants (NPPs) leads to increased electricity generation cost. An in-depth view on spurious activation of digital plant protection systems of NPPs for human errors in maintenance tasks is presented in this paper. A new model which considers human errors in maintenance and periodic tests to predict component failure rates is presented. The model has been applied to OPR-1000 reactor protection system for quantification of spurious trip frequency by fault-tree analysis. The major causes of spurious activation in a nuclear reactor protection system are identified. A set of case studies has been performed with the variation of magnitudes of human errors probability and maintenance strategies, in which, the human errors in maintenance are found to significantly influence reactor spurious trip frequency. This study is expected to provide a useful mean to designers as well as maintainers of the digital reactor protection system to improve plant availability and safety.