核电厂人员闸门新型传动机构研发设计与分析

采用数字化设计方法,研发了人员闸门新型传动机构并完成了样机模块化设计开发。使用故障模式及影响分析（FMEA）方法,找出了新型传动机构薄弱环节,即锁紧机构传动链和地震后的门旋模块及门栓插拔模块,分别利用动力学仿真软件ADAMS对锁紧机构传动链的锁紧功能及运动性能进行了运动模拟仿真分析,利用有限元分析软件ABAQUS对地震后门旋模块的可运行性及门栓插拔模块的有效性进行了分析。结果显示,锁紧机构传动链选型合适,门旋模块满足可运行要求,门栓插拔模块满足有效性要求。研发的新型传动机构传动链及结构设计合理,满足三代核电的要求。      

核电厂人员闸门顶升机构动力学仿真分析

为确保核电厂人员闸门顶升机构结构设计的合理性，对其在密封门启闭过程中的运动特性进行分析研究。本研究借助动力学仿真分析软件（ADAMS）对人员闸门顶升机构可能产生的故障失效模式进行了仿真分析，主要包括顶升机构传动箱运动学分析和传动丝杠动力学分析。分析结果表明，顶升机构传动箱的安装初始位置是导致顶升高度不足的直接因素；通过设置传动箱的初始安装位置、丝杠转速和运行时间可确保顶升高度达到设计要求的60 mm；此外，通过对主要动作部件的受力模拟分析，得出丝杠在径向上承载力可忽略，不会造成丝杠径向断裂事故出现。      

核电厂设备闸门提升装置关键失效模式分析

对核电厂设备闸门提升装置在闸门封头开启和关闭过程中的运动特性进行分析研究,该研究借助动力学仿真软件ADAMS和有限元分析软件ANSYS展开,分别对故障树分析发现的薄弱环节进行分析,包括:提升装置两侧导向箱不同步而产生卡滞和地震工况下闸门封头两侧导轨变形失效。分析结果表明,闸门封头两侧导向箱不同步高度差≤54mm时,提升装置不会产生卡滞现象;在地震工况下,提升装置无论是在闸门封头开启还是关闭状态,两侧导轨都不会出现变形失效。     

设备闸门提升与导向装置运行特性仿真分析

对核电厂设备闸门提升与导向装置在设备闸门封头开启与关闭过程中的运动特性进行了分析研究，该研究借助动力学仿真软件ADAMS展开，分别对提升与导向装置可能出现的同步提升操作和不同步提升操作两种工况进行了仿真分析。分析结果表明:在同步提升操作工况下，设备闸门封头两侧导向箱位移一致，提升与导向装置不会产生卡滞；在不同步提升操作工况下，设备闸门封头两侧导向箱最大允许高度差为54 mm，否则提升与导向装置将产生卡滞。      

基于RAVEN的SGTR事故概率安全裕度分析方法研究

介绍了一种基于RAVEN软件通过蒙特卡洛（MC）抽样的风险指引的安全裕度特性分析（RISMC）方法,综合分析热工参数、人员动作时刻不确定性对蒸汽发生器传热管破裂（SGTR）事故安全裕度的影响,并将计算结果与传统安全评价方法进行比证。针对事故关键影响参数,基于MC抽样量化影响安全裕度的关键参数样本,利用RELAP5程序建立SGTR事故的系统仿真模型,通过RAVEN软件进行耦合计算并加以分析,最终获得该电厂模型在辅助给水系统失效情况下SGTR事故的概率安全裕度及其对各影响参数的敏感度。      

核电厂设备闸门外压极限承载特性研究

采用有限元分析方法对设备闸门外压极限承载特性进行研究。在ANSYS程序中建立参数化计算模型，通过非线性屈曲分析得到设备闸门封头的外压极限载荷。计算结果与经验公式结果相近，验证了有限元分析方法的可靠有效性。对影响外压极限承载特性的设计参数进行了敏感性研究，总结出其影响变化规律。优化设计示例表明该研究方法具有实用性，可为设备闸门的结构设计提供优化方案。      

压水堆核电站运行堆芯物理过程的PC仿真

阐述了PWR核电站堆芯的模型化问题，建立了适用于微机仿真的核电站的临界堆中子动力学模型，温度效应中子动力学模型和堆芯热传递模型。应用所建模型，建立传递函数，用微机仿真并对仿真结果进行分析。     

双层安全壳人员闸门力学分析

在事故工况下,人员闸门需承担防止放射性物质不可接受释放的功能。采用ANSYS有限元分析软件,建立人员闸门的三维模型对三代核电厂双层安全壳人员闸门进行力学分析,并结合土建影响(强迫位移),得到闸门各部分的应力分布以及变形量,以实现安全壳完整性所必须达到的功能。     

基于多学科协同仿真的控制棒驱动机构动力学分析研究

控制棒驱动机构(CRDM)动作过程的计算分析一般涉及多体动力学、电磁学、流体力学等多个学科,并且动作过程中各学科分析所需的输入与彼此计算分析获得结果相互影响,导致动力分析较为困难。使用软件Adams建立了多体动力学模型,使用AnsysMagnetic建立了电磁分析模型,使用AnsysFlotran建立了水阻分析模型,在MatlabSimulink中搭建了CRDM动力学仿真系统,将多体动力学模型、电磁分析模型、水阻分析模型联接,进行协同仿真,仿真获得提升线圈开始通电到提升衔铁碰撞隔磁片的时间为115 ms,CRDM样机试验获得的提升衔铁动作时间为125 ms,仿真结果与试验数据相近,误差为8%。     

基于Simulink输入阶跃反应性时有温度和毒物反馈的反应堆动态响应仿真

计算机仿真对核反应堆运行和核电人员的培训具有十分重要的意义.基于Simulink仿真软件,本工作对点堆中子动力学方程输入阶跃反应性考虑6组缓发中子时具有温度和毒物反馈的特性进行研究,计算了某型反应堆在两种典型工况下引入正、负阶跃反应性时各主要运行参数的变化规律,并将最终的结果与三维模型计算数据进行了对比.结果表明:利用Simulink进行点堆仿真研究能够高效、便捷地满足系统要求,且仿真结果与三维模型计算的数据符合得较好.     

核电厂人员闸门气动回路优化设计与密封检测工艺改进研究

介绍了核电厂人员闸门气动回路设计方案，并结合人员闸门可能出现的失效工况，对气动回路进行了优化设计；从人员闸门存在的整体密封失效和局部密封检测准确性方面，对人员闸门密封检测工艺进行了改进。优化改进后的气动回路设计方案及密封检测工艺已成功应用到在建的三代堆型上，在确保人员闸门人员操作安全性得到大大提升的同时，也有效避免了人员闸门可能存在的密封失效而导致第三道安全屏障完整性受损的事件出现。      

在役核电厂人员闸门传动齿轮优化改进

采用ANSYS有限元分析程序，对国内某在役核电厂人员闸门断齿传动齿轮进行了应力分析，找出了传动齿轮断齿失效的根本原因:传递扭矩过大而引起齿轮过载，应力计算值超过了材料许用应力限值，致使传动齿轮产生根切断齿事故。结合传动齿轮断齿失效根本原因，从材料选择和结构设计2个方面对传动齿轮进行了优化改进。应力分析与评定结果显示，优化改进后的传动齿轮和与之配合的扇形齿轮应力计算值均小于材料许用应力限值，优化改进方案有效降低了传动齿轮的应力水平，有效提高了人员闸门齿轮传动操作的安全性和可靠性。      

在役核电厂人员闸门传动齿轮优化改进

ANSYS finite element analysis program was used to analyze the stresses of the transmission gear for an in-service nuclear power plant in China, and the causes for tooth rupture were found. The transmission torque was so large that the calculated stress exceeded the allowable stress limit of the material, which caused a root rupture accident of the transmission gear. Combined with the causes of tooth rupture for transmission gear, optimization was performed based on material selection and structural improving design. The optimized design scheme passed stress analysis and evaluation. The results show that the calculated stresses of the optimized transmission gear and the sector gear were smaller than the allowable stress limits of the material. The optimized scheme effectively reduced the stress level of the transmission gear, and obviously improved the safety and reliability of the transmission gear operation for personnel airlock.     

数字化核仪表系统样机堆上试验及验证

针对目前国内核电厂核仪表系统设备主要依赖进口的现状,设计研发了一套数字化核仪表系统样机,系统样机主要包括中子探测器组件、信号调理和处理样机以及信号监控设备。通过介绍样机在商用堆上的安装和试验情况,详细分析了反应堆启堆、升功率、满功率及降功率运行期间的试验数据。试验结果表明,中子探测器与信号调理和处理样机配合良好,整套系统样机运行稳定可靠。      

Leakage potential of LWR containment penetrations under severe accident conditions

This paper is an overview of a Sandia National Laboratories, Albuquerque (SNLA) study of the performance of mechanical penetrations in light-water reactor (LWR) containment buildings that are subjected to severe accident environments. The study is concerned with modes of failure as well as the magnitude of leakage. The following tests have been completed, are under way, or are planned: (a) seals and gaskets have been tested to register the effects of radiation aging, thermal aging, seal geometry, and squeeze on seal and gasket materials in severe accident environments; (b) the performance of a full-scale airlock will be evaluated at severe accident temperature and pressure levels; (c) personnel airlock and equipment hatch tests were made on a model of a steel containment building; and (d) tests of mechanical penetrations are planned as part of a test on a model of a reinforced concrete building. This program is part of an overall US Nuclear Regulatory Commission (USNRC) effort to evaluate the integrity of LWR containment buildings.     

Quantitative modeling of digital reactor protection system using Markov state-transition model

Recently, digital instrumentation and control systems have been increasingly installed for important safety functions in nuclear power plants such as the reactor protection system (RPS) and the actuation system of the engineered safety features. Since digital devices consist of not only electronic hardware but also software that can control microprocessors, the functions specific to digital equipment such as self-diagnostic functions have been becoming available. These functions were not realized with conventional electric components. On the other hand, it has been found that it is difficult to model the digital equipment reliability in probabilistic risk assessment (PRA) using conventional fault tree analysis technique. OECD/NEA CSNI Working Group of Risk Assessment (WGRisk) set up the task group DIGREL to develop the basis of reliability analysis method of the digital safety system and is now discussing about several issues including quantitative dynamic modeling. This paper shows that, taking account of the relationship among the RPS failures, demand after the initiating event, detection of the RPS fault by self-diagnostic or surveillance tests, repair of the RPS components and plant shutdown operation by the plant operators as a stochastic process, the anticipated transient without scram (ATWS) event can be modeled by the event logic fault tree and Markov state-transition diagrams assuming the hypothetical 1-out-of-2 digital RPS.     

Failure Modes and Reliability Analysis of Motor Driving System in ITER Remote Handling Transfer Cask

For remote handling in a far distance and toxic substances leakage must be considered, International Thermonuclear Experimental Reactor has put forward higher requirement for the failure modes and reliability analysis of remote handling system. Based on the actual working condition of motor driving system when transfer cask remote handling the in-vessel components, all the failure modes of the motor driving system were analyzed and distinguished by software Isogragh, the main failure patterns and weak links were the control circuit failure mode, bearing failure mode, electrolytic capacitor and the insulation failure mode. Motor driving system reliability model in series was established, the reliability and mean time between failures were estimated by the component stress analysis method. The above-mentioned methods and results provide the basic references for the system reliability assessment, optimization design, making rapid and accurate judgment for potential failure when systems operating.     

Performance of containment penetrations under severe accident loadings

The paper provides a summary of efforts to date to better understand the leakage behavior of containment penetrations when subjected to severe accident conditions. The research activities discussed herein are a part of the Containment Integrity Programs, which are managed by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. Past containment penetration research topics, which are briefly described, include testing of typical compression seals and gaskets, electrical penetration assemblies, and a personnel airlock, as well as an investigation of leakage due to ovalization of penetration sleeves. The primary focus of the paper is on recent or ongoing research programs on the behavior of inflatable seals, bellows, and of pressure unseating equipment hatches.