数字化人机界面人因可靠性分析方法研究与应用

为了将人因工程方面有关人的能力和限制的知识应用到人机界面的设计,从而使控制室系统设计达到人‐机‐环境的最佳匹配,本文研究通过人因可靠性分析方法,结合人因工程设计过程,建立一种适于工程应用的综合性分析方法来识别人机界面中影响人员绩效和容易诱发人因失误的潜在设计缺陷,并采用系统化的方法来优化人机界面设计。结果表明,本文建立的方法具有可操作性强、评价客观等优点,可有效提高核电厂安全性、可靠性和经济性。该方法现已成功应用于在建的CPR1000各项目,具有广阔的应用空间。     

临界装置数字化人机界面的分析与实现

为了建立信息化程度高、扩展灵活、适合操纵人员需要的全新数字化系统,将数字化人机界面（HMI）技术应用于临界装置控制系统中。通过对临界装置数字化人机界面的分析,与人因工程原则相适应,描述了符合临界装置运行习惯和任务处理特点的数字化运行理念,最终实现了临界装置的数字化控制监测保护系统,并投入稳定运行。     

基于人因可靠性的核电厂数字化人机界面功能布局优化方法研究

监视信息过程中,人的可靠性很大程度上取决于数字化人机界面中功能块之间的布局。本文提出了基于人因可靠性的核电厂数字化人机界面功能布局优化方法,对该优化方法建立了一个完整的优化流程,提出了线性逆向杂交方法,使用人因可靠性作为优化标准。对该优化方法进行实验分析,实验结果表明,线性逆向杂交方法有较好的稳定性,本文提出的优化方法有较好的精确性及收敛性。     

人因工程与控制室

本文主要介绍了秦山第二核电厂控制室（包括主控室和辅助控制点，下同）在设计过程中贯彻人因工程适用原则的情况。并简要说明在核电站控制室的设计中贯彻人因工程原则的重要性。     

核电厂人因工程人机接口设计的探讨

本文论述了人因工程在核电厂设计中的重要性。简要介绍了人因工程的内容和范围,重点讨论了新建核电厂的人机接口设计。简要说明了国外核电厂人机接口设计,详细描述了国内核电厂的人机接口设计要求。     

核电厂控制室的人因工程

核电厂控制室是核电厂人－机接口最集中的地方，也是核电厂人误发生率最高的地方。发展先进的核电厂控制室是提高核电厂安全性和可靠性最有效的方法。先进控制室的发展。除了不断引入先进的技术和设备之外，最根本的是考虑了人因工程的原则，在简化操作减轻运行人员工作负相的同时，大大提高了核电厂的安全性和可靠性     

核电站计算机化规程设计的人因工程要求

在复杂的安全重要系统中,需使用规程支持人员任务,以减少失误并提高系统安全。在核电行业,随着数字化技术的发展,许多计算机化规程系统得到了开发和应用,可以支持操纵员进行信息搜集、判断决策和规程执行。在此基础上,IEEE 1786、IEC 62646和NUREG—0700等相关的人因工程标准导则对计算机化规程的设计和评审进行了规范。本研究调研了计算机化规程系统的设计和应用情况,从规程的呈现、规程步骤的标记、规程的管理、系统参数显示、参数监视与任务支持以及系统控制等方面进行了总结。在此基础上综述了相关标准导则的条款,并对计算机化规程的设计要求和一般设计原则进行了讨论。本研究希望通过总结相关信息,探讨计算机化规程人机接口设计的实践经验和人因工程要求,为设计和评估人员提供参考。     

高能量工程试验堆主控室的人因工程应用

应用人因工程原理对高通量工程试验堆（ＨＦＥＴＲ）主控室人－机接口，仪器设备及工作环境进行了技术改造；参照国际惯例，增设了反应堆安全参数显示系统，改善、加强了人、机的相互作用，提高了ＨＦＥＴＲ的安全性。     

高通量工程试验堆主控室的人因工程应用

简要说明了高通量工程试验堆（HFETR）的人—机体系;介绍了应用人因工程的原理改善 HFETR 主控室工作环境,提高设备、仪器的性能和操作的可靠性;论述了 HFETR 人因失误的减少与人因工程的关系。提出了进一步加强人因工程的应用,可以提高主控室人—机系统的交互作用效果,并能够为反应堆更加安全、可靠地运行提供条件保证。     

高通量工程试验堆主控室的人因工程应用

应用人因工程原理对高通量工程试验堆（ＨＦＥＴＲ）主控室的人机接口，仪器设备及工作环境进行了技术改造；参照国际惯例，增设了反应堆安全参数显示系统。改善、加强了人机的相互作用，提高了ＨＦＥＴＲ的安全性。     

Overview of engineering design,manufacturing and assembly of JT-60SA machine

The JT-60SA experiment is one of the three projects to be undertaken in Japan as part of the Broader Approach Agreement, conducted jointly by Europe and Japan, and complementing the construction of ITER in Europe. The JT-60SA device is a fully superconducting tokamak capable of confining break-even equivalent deuterium plasmas with equilibria covering high plasma shaping with a low aspect ratio at a maximum plasma current of I_p = 5.5 MA. This makes JT-60SA capable to support and complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction. After a complex start-up phase due to the necessity to carry out a re-baselining effort with the purpose to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility, in 2009 detailed design could start. With the majority of time-critical industrial contracts in place, in 2012, it was possible to establish a credible time plan, and now, the project is progressing on schedule towards the first plasma in March 2019. After careful and focused R&D and qualification tests, the procurement of the major components and plant is now well advanced in manufacturing design and/or fabrication. In the meantime the disassembly of the JT-60U machine has been completed and the engineering of the JT-60SA assembly process has been developed. The actual assembly of JT-60SA started in January 2013 with the installation of the cryostat base. The paper gives an overview of the present status of the engineering design, manufacturing and assembly of the JT-60SA machine.     

Evaluation of human factors in interface design in main control rooms

An evaluation of human factors in a new nuclear power plant was conducted prior to the beginning of any business operations. After the task analysis and observation of training, two stages of interviews were carried out with the operators in the Fourth Nuclear Power Plant (NPP4). The main concerns identified were problems resulting from the operating interface of the display and controls in the main control room, usability of procedures, and the layout of the main control room.The latent human errors and suggestions were listed, and the top three problems were analyzed. The operators indicated that the alarm design issues and the critical problem of the operating mode with the VDU were worth further study in order to provide suggestions for a new interface design for future power plants.     

基于USB通信的多道分析器接口设计

介绍了一种新型多道分析器的接口电路设计，它采用USB技术与计算机进行通信，使多道分析器具有了通用性强、即插即用的优点；同时接口中采用了双口SRAM和高速MCU，用于数据的存储和传输，解决了接口向计算机传输数据的瓶颈问题。     

基于Oracle的SSRF数据存档系统及接口设计

上海光源(SSRF)采用EPICS作分布式控制系统的软件平台。Channel Archiver是EPICS自带的OPI层数据存档工具集,实现对EPICS控制系统运行过程中各种数据的快速存取及检索。本文对EPICS数据存档系统进行分析,提出现有的不足,并针对这些不足,结合Oracle数据库对存档系统重新设计,实现中央数据库统一存储和管理加速器运行数据,并采用Web Services技术提供访问数据库的统一接口。     

数控核测井仪中的实时稳谱程序设计

详细地讨论了实现数字四能窗稳谱技术的算法程序，结合γ射线测井仪，从原理，方法，测井实际需求等方面进行了较深入的分析，研究和实际检测，提出了较完整平时稳谱控制流程和程序实现方法，并在γ射线测井中得到了较好的验证。     

用于非接触式核测量的红外通信接口的设计

介绍了一种简易、低成本的PC机与单片机进行红外通信方式,并介绍了相应的接口电路。该方法可用于多机通信。     

地下实验室主体结构方案设计影响因素分析

主体结构形式的确定是高放废物处置地下实验室工程设计的关键内容。从地下实验室类型、功能要求、深度确定、用地条件、地形条件、工程地质条件、建造实施、运行安全、长期安全性及许可证申领、环境保护和工期与造价等方面讨论和分析对影响主体结构方案设计的因素,并提出了设计建议,以期为我国地下实验室总体方案设计提供参考。     

Basic design guideline for the preliminary engineering design of PIE facilities in IFMIF/EVEDA

The international fusion materials irradiation facility (IFMIF) is an accelerator-based intense 14 MeV neutron source for testing fusion reactor materials. Under broader approach (BA) agreement between EURATOM and Japan, the engineering validation and engineering design activity (EVEDA) were started from 2007. The IFMIF needs the post irradiation examination (PIE) facilities to generate a materials irradiation database for the design and licensing of fusion DEMO reactors. In this study we examined and discussed about the safety such as remote handling, hot cell design, and the equipments and apparatus of hot cells, and we summarized a basic design guideline for the preliminary engineering design of the PIE facilities.     

A computerized method for flow-induced random vibration analysis of nuclear reactor internals

The general analytical, numberical, and programming techniques of a computerized method for flow-induced random vibration analysis of nuclear reactor internal components is discussed. The statistical approach used is similar to that originally introduced by Powell and subsequently applied to predict the response of flat plates to homogeneous turbulent air flow. The input damping ratios and parameters related to the flow field are assumed to be known from experimental data, while the virtual mass and natural frequency shift effects due to hydraulic loading of the structure are included in the analysis. The latest numerical techniques developed for use with modern, high-speed digital computers are employed to evaluate the acceptance integrals, thus permitting the basic method to be applied to the vibration analysis of complex structures excited by inhomogeneous turbulent flow — a situation that is commonly encountered inside a nuclear reactor. The importance of computer program modulization and its relationship to overlays are discussed. Some representative predicted vibration amplitudes based on a typical pressurized water reactor design are given.     

Main R&D issues for fast reactor structural design standard

For realization of economical and reliable fast reactor (FR) plants, the Japan Atomic Energy Agency (JAEA) and the Japan Atomic Power Company (JAPC) are cooperating on the “Feasibility Study on Commercialized FR Cycle Systems”. To certify the design concepts through evaluation of the structural integrity of FR plants, the research and development of the “Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)” is recognized as an essential theme. The FDS focuses on particular failure modes of FRs such as ratchet deformation and creep-fatigue damage due to cyclic thermal loads. For precise evaluation of these modes, the research and development for three main issues is in progress. First, the “Refinement of Failure Criteria” needs to be addressed for particular failure modes of FRs. Secondly, the development of “Guidelines for Inelastic Design Analysis” is conducted to predict elastic plastic and creep deformation under elevated temperature conditions. Lastly, efforts are being made toward preparing “Guidelines for Thermal Load Modeling” for the design of FR components where thermal loads are dominant.