秦山第二核电厂安全壳结构在线监测系统改造及应用

利用基于激光测量技术的测微传感器实现安全壳水平变位监测系统改造,利用具有模拟信号输出功能的精密位移传感器替代安全壳垂直变位监测装置,再将安全壳变位、混凝土应变、灌油钢束预应力、混凝土温度等测量参数全部传输至系统监控主机,实现整个系统数据的自动采集、传输、存储和处理。改造后的安全壳结构在线监测系统优点突出,在秦山第二核电厂安全壳整体结构试验及日常安全壳结构性能监测中得到了较好的应用,可为其他同类型核电机组提供借鉴。     

“华龙一号”非能动安全壳热量导出系统布置优化研究

非能动安全壳热量导出系统(PCS)是我国自主设计的具有完整知识产权的第三代核电技术“华龙一号”中非常重要的一个非能动的安全系统,其通过布置在安全壳外部的换热水箱将安全壳内的热量导出到壳外的大气环境中,以这种自然循环的运行实现降低安全壳内温度和压力的目的。     

基于Wi-Fi的远距离无线数据采集系统

传统的军用摩托车测试系统大多是存储测试系统,无法将测试数据实时地传递到远端的控制台,这给数据的读取和分析带来不便.论文设计并实现了一种基于Wi - Fi远距离无线数据采集系统.相比目前其它无线传输系统该无线采集系统具有传输速度快、距离远、抗干扰能力强、能与以太网整合等优点.在满足传输速率的前提下,通信距离可达10 km.     

多级蓝牙无线数据采集传输系统

提出一种基于单片机、多级蓝牙模块的高速、高旋飞行弹体无线数据采集传输系统,系统以多级蓝牙模块无线传输系统为核心,通过软硬件实现炮弹发射前初始参数、弹体出炮口初速等数据采集、发送、接收、处理的目的。实践证明,多级蓝牙无线数据传输系统在恶劣电磁屏蔽和高速旋转环境中可以可靠传输数据,为各种型号火炮初始、炮口初速等参数的装订、传输提供了应用依据[1]。     

GEM探测器高速数据采集系统设计

介绍了基于以太网的GEM探测器高速数据采集系统的设计。该系统将GEM探测器输出的电荷信号转换为数字信号并写入FPGA进行分析和处理,处理后的数据通过千兆以太网进行传输。主机电脑接收以太网传输的电荷信号的位置信息,绘制电荷信号的位置分布图。实验测试表明:该系统能检测到探测器输出的位置信息并绘制出X射线信号的位置分布图。     

高压电力电缆接头温度无线检测控制系统设计

基于无线传输技术、单片机技术、虚拟仪器技术等设计了一种电力电缆中间接头故障检测控制系统,系统利用低功耗数字温度传感器TMP100实现地下电缆接头温度采集,单片机MSP430F149作为中央核心控制单元完成数据分析与处理,并通过nRF401无线收发器实现了数据无线传输.文章对系统部分硬件及软件设计进行了详细介绍.通过实验测试,系统较好地满足了设计要求.     

地下核电站安全壳再循环系统设计的初步论证

核电站建造于地下,反应堆厂房洞室外具备天然的裂变产物屏障,在安全壳外洞室内设置安全壳再循环系统,预防并缓解放射性裂变产物释放,维持安全壳的完整性。该系统同时整合了卸压、过滤、排热安全功能,充分发挥地下核电站重力补水和天然屏障的安全优势,可以非能动运行。本文通过简单的计算分析开展初步论证,证明该系统可以有效实现三大安全功能,是适合于地下核电站的安全系统。     

车辆防撞监控预警系统研究与设计

随着现代社会程度的高度发展,交通安全问题已成为各国亟待解决的重大课题.论文提出了一种利用超声波技术实现防撞监控预警系统.该系统利用超声波传感器实现距离信号采集与接收,通过单片机AT89C52实现信号分析、处理及声光报警、减速及刹车制动等操作.为保证系统测距精度,利用数字温度传感器DS18B20实现超声波传输速度补偿;为使系统智能性更强,利用红外光电传感器实现测速;利用行程开关检测车辆所处直行、转弯或超车等状态,保证车辆在不同状态下,安全距离的准确计算及预警.文章详细介绍了系统设计原理及部分软硬件设计方法.     

核电站安全壳结构整体性能检测标准与技术

1 前言 安全壳是核电站反应堆的最后一道安全屏障,对核电的安全至关重要。根据国际原子能机构为规定和国际惯例,核电站建成后,必须经过安全壳结构整体性试验（SIT）,检验安全壳在构造、强度和施工质量方面承受失水事故工况的能力。检测评定合格,方能装料发电。 安全壳结构检测项目（SIT）,除测试传感器、数据采集处理等试验技术外,还包括安全壳结构分析,实测与计算的吻合分析、安全评估等多项工作内容。 压水反应堆核电厂的安全壳有钢结构、钢筋混凝土结构、预应力混凝土结构几种形式,其中预应力混凝土结构由于性能好,近年来得到各核电国的重视。美、日、法等国家对该种安全壳,从原材料、节点构造到施工工艺、模型试验等进行过系统的试验研究。对于安全壳结构整体性试验（SIT）,也建立起一套较为完整的测试系统和技术制度,编制了相应的规程和标准。     

一种用于旋转设备运行状态无线监测的智能无线振动传感器设计

为提高已投入运行核动力装置旋转设备的运行数据采集和状态监测能力,需要解决安装传感器和敷设配套线缆困难的问题。本文采用现场可编程门阵列（FPGA）作为主控单元,设计了一种基于Zigbee物联网通信技术的智能无线振动传感器,并给出了其电路构成、工作原理,以及嵌入式控制软件的工作流程。通过对此传感器进行性能测试,结果表明该传感器功耗低,实现了对振动信号的连续采集、智能分析与上传。该无线传感器安装简单,无需敷设供电和信号线缆,可应用于构建核动力装置旋转设备的状态监测系统。      

Scaling of passive condenser system for separate effect testing

The passive containment cooling system (PCCS) of the simplified boiling water reactor (SBWR) is a passive condenser system designed to remove energy from the containment for long term cooling period after a postulated reactor accident. Depending on pressure condition and noncondensable (NC) gas fraction in drywell (DW) and suppression pool (SP), three different modes are possible in the PCCS operation namely the forced flow, cyclic venting and complete condensation modes. The prototype SBWR has total of six condenser units with each unit consisting of hundreds of condenser tubes. Simulation of such prototype system is very expensive and complex. Hence a scaling analysis is used in designing an experimental model for the prototype PCCS condenser system. The motive for scaling is to achieve a homologous relationship between an experiment and the prototype which it represents. A scaling method for separate effect test facility is first presented. The design of the scaled test facility for PCCS condenser is then given. Data from the test facility are presented and scaling approach to relate the scaled test facility data to prototype is discussed.     

压水堆核电站安全注入试验期间执行机构拒动和误动的干预对策分析

安全注入试验是压水堆核电厂热试期间涉及范围最广、风险最高的试验。试验程序要求在热停平台通过快速开启蒸汽排放阀模拟二回路破口触发安注信号,验证反应堆跳闸,安全壳隔离,安注执行机构动作,并对开盖冷试期间调整的安注流量进行再次验证。安注信号一旦触发将导致22个系统共计234个设备真实动作,一回路被注入含硼水。任何在线错误、设备缺陷或操作失误都可能导致试验失败,甚至可能导致一回路设备损坏;同时因安全注入试验将导致核电站主回路产生一次瞬态,对一回路设备冲击极大,所以安全注入试验必须保证一次成功。为了保证试验的真实性及完整性,提高试验的一次成功率,控制试验的风险,本研究针对以往项目执行该试验时存在的一回路水位过高及设备误动或拒动的难题,对试验方案进行了优化创新。该方案成功运用于阳江3号机安全注入试验,一定程度上解决了稳压器水位过高及设备误动、拒动的难题,获得了机组安全可控且试验顺利高效的效果,达到了同行领先水平。     

Analysis of maximum pressure in VVER1000/V446 reactor containment for LOCA and MSLB

The highest thermal-hydraulic pressure in the containment occurs when reactor coolant in the first loop and steam in the secondary loop discharge simultaneously,and when the maximum amount of energy from reactor unit enters to containment volume.In this paper,we investigate temperature and pressure variations in the VVER1000 containment compartments owing to concurrent break in the pipelines of the primary and secondary loops.A two-phase,multicellular model is applied in the presence of non-condensable gases.Convection and conduction through the main heat structures inside the containment are also considered.The predicted results agree well with available data.Maximum values of pressure and temperature in the containment are then calculated and compared to the design values.If LOCA and MSLB occur simultaneously,the maximum pressure would exceed the design value and integrity of the containment would be threatened.     

Gabor变换用于核压力容器裂纹楞边超声检测

核压力容器不锈钢堆焊层的材料噪声降低了在役检测中超声信号的信噪比。本文应用Ｇａｂｏｒ变换时频分析技术，根据裂纹楞边的频偏特性，提取了裂纹楞边超声回波。该技术可用于定量地检测核压力容器的裂纹状况。     

Prototype EMAT system for tube inspection with guided ultrasonic waves

A prototype system with full computer support for ultrasonic inspection of ferritic tubes using guided waves is described. The ultrasonic waves are launched and received with the aid of electromagnetic acoustic transducers which are layed out as linear phased arrays. The array structure provides a good axial directivity for the transducers so that the probe can be positioned anywhere along the tube length sequentially transmitting ultrasonic pulses in the foreward and backward directions. While the probe is fixed at one axial position during inspection the tube length is measured by the system and flaws are detected from returning ultrasonic echos. Results of the inspection of tubes with natural flaws are given and the wavelength-spectrum of the ultrasonic mode used for the inspection is discussed with respect to flaw depth sizing.     

Coolant cavitation in dynamic containment loading

After a brief introduction to the subject of cavitation in subcooled liquids and a survey of what is known regarding the key parameters in the cavitation process for water and for sodium, the basic equations of the SIMON cavitation model for use with Lagrangian containment codes and the assumptions behind them are reviewed.Some calculations using this model are then presented which show the dissipative effect of cavitation both in uncavitated liquids transmitting tension waves and in cavitated liquids transmitting pressure waves. The cavitation which develops when pressure waves are reflected at free surfaces is also examined, and some calculated results are compared with an experiment involving this phenomenon found in the literature. The role of cavitation in the containment loading process is then discussed, and examples taken from model test calculations are adduced to show that cavitation occurs at all stages of the loading process and involves a high proportion of the total liquid volume. Again by example the point is made that in certain simple circumstances a crude pressure cut-off model of cavitation is adequate but that for other major aspects of the containment loading process such as roof impact pressures and structural deformations a more refined model is necessary.     

核电厂安全壳首次整体在役试验方案优化

安全壳整体试验是压水堆核电机组一项特大型、高风险、高难度的试验,通过模拟设计基准事故工况下安全壳内的峰值压力,在事故峰值压力平台下,进行安全壳整体泄漏率测量及各压力平台安全壳结构试验,以验证其密封和结构性能。安全壳整体试验是国家核安全局监管的一个重要见证点,试验结果直接决定是否能够启动反应堆发电。301大修安全壳整体试验是3号机组首次在役试验,本次试验汲取了秦山第二核电厂以往6次安全壳整体试验的经验和其他电厂的反馈,试验方案更加科学,试验的组织管理更为规范。文章对301大修安全壳整体试验的经验进行了论述和总结,希望对电厂以后的安全壳整体试验提供参考。     

Leak-tightness characteristics concerning the containment structures of the HTTR

The containment structures of the HTTR consist of the reactor containment vessel, the service area, and the emergency air purification system, which minimise the release of fission products in postulated accidents, which lead to fission product release from the reactor facilities. The reactor containment vessel is designed to withstand the temperature and pressure transients and to be leak-tight in the case of a rupture of the primary concentric hot-gas duct, etc. The pressure inside the service area is maintained at a negative pressure by the emergency air purification system. The emergency air purification system will also remove airborne radioactivity and will maintain a correct pressure in the service area.The leak-tightness characteristics of the containment structures are described in this paper. The measured leakage rates of the reactor containment vessel were enough less than the specified leakage limit of 0.1%/d confirmed during the commissioning tests and annual inspections. The service area was kept in a way that the design pressure becomes well below its allowable limitation by the emergency air purification system, which filters efficiency of particle removal and iodine removal well over the limited values.The obtained data demonstrate that the reactor containment structures were fabricated to minimise the release of fission products in the postulated accidents with fission product release from the reactor facilities.     

Estimation of ex-vessel steam explosion pressure loads

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment.In this article, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel–coolant interactions. A parametric study was performed varying the location of the melt release (central, right and left side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to establish the influence of the varied parameters on the fuel–coolant interaction behaviour, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. For the most explosive central, right side and left side melt pour scenarios a detailed analysis of the explosion simulation results was performed. The study shows that for some ex-vessel steam explosion scenarios higher pressure loads are predicted than obtained in the OECD programme SERENA phase 1.     

非能动安全壳冷却系统空气流道自然循环的比例分析与设计

在考虑建设试验台架经济性的前提下,缩小比例的单项和整体效应试验台架对研究和开发大型先进压水堆核电站及其分析验证程序都具有重要意义。非能动安全壳冷却系统（PCS）壳外空气流道内的自然循环在安全壳非能动冷却性能中发挥着重要的作用。本文从自然循环的数学模型出发,推导出了单项和整体效应试验台架的比例设计方法。在给定壳内热流密度的条件下,通过PCCSAP-3D程序对CAP1400非能动安全壳的2/5比例单项效应试验理想比例台架（ISF）进行模拟。结果表明,本比例分析与设计方法以及在降低高度台架上模拟自然循环是可行的。