Computerized procedure system for the APR1400 simulator

The ShinKori Unit 3&4 which is the reference plant for APR1400 simulator plans to utilize various display devices and computerized Human Machine Interface (HMI) as well as many kinds of digitalized instrumentation and control. The operating crew will be given a lot of information on display devices in order to diagnose the system states and control devices. And with the limitations of paper-based procedure (PBP) including lack of interactive capabilities between the PBP and sophisticated systems, and the high volume and cost associated with paper manuals used in complex and technically demanding environments, the computerized procedure system (CPS) is developed as a part of computerized systems for APR1400 simulator. The developed CPS covers all operating procedures including normal, abnormal and emergency, and provides information through workstation consoles using flowchart and tree structure and also has automatic checking capabilities of step logic including continuously applied steps. The CPS consists of Procedure eXecution System (PXS), and Engineering System (ES) including procedure editor (PED) and Non-Integrated System Test System (NISTS). The PXS is the execution system for the procedure provided to the operators and has the following functions. 1) Display of the plant status information necessary for the instructions and executions for the procedure, 2) auto-check functions instead of manual monitoring based on the paper-based procedures, 3) detection and recovery functions of operators’ faults, 4) generation function of the LCP (Logged Computerized Procedure) to search the results of the procedure execution, 5) generation function of the Log Paper Procedure supporting the continued operation using the paper procedure when the CPS fails. The ES (Engineering System) is the environment provided to the writer and maintainer of the CP (Computerized Procedure), the application program loaded in the PXS. The CP is frequently revised in the plant, and related to the safety of the plant. Therefore, the CP is provided to secure the quality and convenience of the developed procedure. For the verification, among the implemented computerized procedures by using PED, a procedure for the LOCA (Loss-Of-Coolant Accident) is selected for testing the integrity of the CPS including communication between CPS and other systems for APR1400 simulator, which is proved to be well implemented. Currently, the implemented CPS is being used to the Preliminary Validation Phase II Test hosted by NETEC (Nuclear Environment TEChnology Institute) in Korea.     

Characterization of an ASIC for CPG sensors with grid-only depth of interaction sensing

A new version of the front-end application specific integrated circuit (ASIC) for co-planar-grid (CPG) sensors is presented. Compared to the first version, the ASIC is optimized for signal/noise performance, modified in peaking time, and it implements timing signals to measure the depth of interaction using the grids signals only. Experimental results using state-of the-art CPG sensors and different approaches for reducing the error due to electron trapping are reported. A new technique that makes use of the sum and difference of the grids signals is also presented.     

A mixed implicit/explicit procedure for soil-structure interaction

This paper describes an efficient method for the solution of dynamic soil-structure interaction problems. The method which combines implicit and explicit time integration procedures is ideally suited to problems in which the structure is considered linear and the soil non-linear. The equations relating to the linear structures are integrated using an unconditionally stable implicit scheme while the non-linear soil is treated explicitly. The explicit method is ideally suited to non-linear calculations as there is no need for iterative techniques. The structural equations can also be integrated explicitly, but this generally requires a time step that is much smaller than that for the soil. By using an unconditionally stable implicit algorithm for the structure, the complete analysis can be performed using the time step for the soil. The proposed procedure leads to economical solutions with the soil non-linearities handled accurately and efficiently.     

数据驱动的核电厂人因失误因果机理模型研究

为了辨识人因失误机理,采用组织定向的人因失误分析（OTHEA）技术对2010—2017年国内核电厂发生的137件人因事件报告进行分析,采用相关性和因子分析方法,辨识人因失误影响因素之间的相关关系,识别出触发人因失误的场景,在此基础上,建立人因失误因果机理模型,揭示人因失误机理。结果表明,影响人因失误主要的结合模式为：知识经验水平、信息显示质量、压力水平、注意力与警觉性、安全态度。知识经验水平主要受培训水平和交流水平的影响;信息显示质量主要受技术系统、人-机界面、规程、组织设计的影响;压力水平主要受任务、规程、人-机界面、技术系统、组织设计以及知识经验水平和信息显示质量的影响;注意力与警觉性主要受工作环境、工作组织管理以及信息显示质量的影响;安全态度主要受组织安全文化、工作组织管理以及操纵员的素质与能力的影响。上述研究可为核电厂人因失误的精准防控提供理论依据,提升核电厂的安全水平。     

“华龙一号”征兆导向应急事故规程开发

In order to compensate for the defects of event-oriented emergency procedure (EOP) and state-oriented emergency procedure (SOP), HPR1000 nuclear power technology takes the advantages of the two operation procedures. Considering probabilistic safety analysis (PSA), a new symptom based emergency operating procedures (SEOP) through a large number of operation analysis supporting calculations is established. As an example, the operator actions during steam line break accident guided by SEOP is studied and compared with EOP and SOP. The results show that SEOP can deal with the accident rapidly and directly and can defend multi-accidents. The accident identification and mitigation measures are reasonable and effective. It can make full use of HPR1000 active and passive safety systems to deal with accidents, give full play to the design advantages of the safety system, and enhance the safety level of HPR1000. The principle, methodology and technique of the development can be used in the procedure development for the similar plant and can be used as a reference to improve the procedures for nuclear power plants in service.     

“华龙一号”征兆导向应急事故规程开发

为了弥补事故导向应急事故规程（EOP）和状态导向应急事故规程（SOP）的缺陷，“华龙一号”核电技术将两者优势相结合。借鉴概率安全分析（PSA），通过大量的运行分析支持性计算，形成全新的征兆导向应急事故规程（SEOP）。以主蒸汽管道破裂事故为例，进行了SEOP引导下的典型事故应用研究及其与EOP和SOP的对比。结果表明，SEOP具有迅速直接处理事故以及较强的叠加事故应对能力，事故判断和缓解措施有效、可靠，能够合理调用能动加非能动安全系统应对事故，充分发挥了“华龙一号”安全系统设计优势，进一步提升了“华龙一号”的安全水平。SEOP开发过程所形成的思路、方法、技术体系，可用于同类核电厂的事故应急规程开发，并可为现役核电厂规程的改进提供借鉴。      

计算机化规程系统在APIO00中的应用

数字化仪控技术在核电厂的推广,为计算机化规程系统（CPS）在以APl000为代表的三代先进压水堆中的应用提供了技术基础。本文对计算机化规程系统的结构、功能和特点进行了描述,CPS作为先进的规程系统,不仅将运行规程以电子化的形式显示在电脑屏幕上,而且同时能监测和处理电厂参数,为操纵员提供规程执行指导,并且显著提高了规程的执行效率。另外,本文还将APl000中的CPS与某M310;0H改进型核电厂的数字化规程系统进行了比较,并结合实际CPS的本土化以及存在问题进行探讨。     

基于本体的核电运行规程模型构建方法

Nuclear power operating procedures are the operational procedures that must be followed for the safe operation of nuclear power plants. Having a lot of texts, the operation procedures are inconvenient for users to browse. However, the current electronic procedures have problems such as lack of simplification, weak content relevance, and less superiority than paper procedures. It is imperative to design a simple and efficient electronic procedure system based Ontology model for nuclear power operation. This paper takes the operating procedures of the chemical and volume systems in a nuclear power plant as an example, uses the ontology model theory, analyzes the text of procedures, develops the knowledge of the procedures and extracts key concepts. Using the ontology development tool to build the system, the electronic procedures and visualization are realized. This construction method is also applicable to the operating procedures of other systems, which not only enables the reuse and sharing of ontology knowledge, but also facilitates querying, understanding, and memorizing procedures. It can help the operators in nuclear power plants to use procedures, thereby improving the efficiency and enhancing the safety and intelligence of nuclear power plants.     

基于本体的核电运行规程模型构建方法

运行规程是核电厂安全运行必须遵守的操作程序。核电厂运行规程繁多、查找不便,现行电子规程存在精简程度不够、内容关联程度弱、相比纸质规程优越性不强等问题。为核电厂设计一种基于本体模型的简洁高效电子化规程系统势在必行。以某核电厂化学和容积控制系统运行规程为例,运用本体模型理论,分析规程文本、挖掘规程知识、提取关键概念。使用本体开发工具构建系统,实现了电子规程及其可视化,可浏览电子规程和知识库,查询关键操作。此构建方法同样适用于其他系统的运行规程,不仅使本体知识实现重用和共享,而且有利于查询、理解、记忆规程,能够为核电厂操纵员使用规程提供帮助,提高效率,增强核电厂的安全性和智能化。      

Guidelines for computerized presentation of emergency operating procedures

New methods of information presentation and interface design have changed the work conditions in the modern nuclear power plant control room. One area that is receiving considerable attention is that of procedures—in particular, emergency operating procedures (EOPs). Such procedures are traditionally available as printed documents; however, developments in information technology have created new opportunities for procedure presentation, and this paper reports on the development of human factor guidelines for computerized presentation of EOPs. After identifying the principle stages in the transition from procedures as documents to fully automated procedures, computerized procedure presentation (CPP) is briefly discussed. Guidelines for the presentation of such are outlined, starting with the high level goals that also constitute the criteria against which the computerized procedures are measured during implementation. Six dimensions that describe CPP are presented and explored in detail, by identifying points in each dimension that characterize different levels of information technology sophistication.     

Computerised Modelling of Radiolysis and Corrosion Effects in Transport Containers for Fuel Spent Nuclear

Abstract

The radiolysis of water and/or gases within transport containers for spent nuclear fuel may result in the generation of hydrogen and oxygen gases and also the enhanced corrosion of the materials in contact with the water. These effects are important, particularly when the fuel container is also used for storage post-transport prior to reprocessing or disposal. The behaviour of a range of radiolytic systems has been studied. Plant behaviour has been simulated in numerous laboratory experiments: plant and experimental results have been linked by a computerised model describing the radiolysis mechanism and predicting the quantities and production rate of gaseous and corrosive species. This allows prediction of plant performance over a long time scale. The model is based on a well-accepted radiolysis mechanism supplemented with specific measurements made at the Harwell laboratory. Model capabilities include inert atmospheres, materials corrosion, variations in water and gas volumes or aqueous chemistry. The model has been applied to design stage radiolysis assessments of transport containers; information from operating plant has been interpreted to advise on design improvement, e.g. diminution of gas production using easily corroded scavengers to remove oxygen. Radiolysis in gas filled dry storage containers for spent nuclear fuel has been studied; corrosive product production (e.g. nitric acid), which is important for fuel cladding integrity has been assessed. The development and use of this computerised model is described with a current summary.     

Integration of functional reliability analysis with hardware reliability: An application to safety grade decay heat removal system of Indian 500 MWe PFBR

A passive system can fail either due to classical mechanical failure of components, referred to as hardware failure, or due to the failure of physical phenomena to fulfill the intended function, referred to as functional failure. In this paper a methodology is discussed for the integration of these two kinds of unreliability and applied to evaluate the integrated failure probability of the passive decay heat removal system of Indian 500 MWe prototype fast breeder reactor (PFBR). The probability of occurrence of various system hardware configurations is evaluated using the fault tree method and functional failure probabilities on the corresponding configurations are determined based on the overall approach reported in the reliability methods for passive system (RMPS) project. The variation of functional reliability with time, which is coupled to the probability of occurrence of various hardware system configurations is studied and incorporated in the integrated reliability analysis. It is observed that this consideration of the dependence of functional reliability on time will give significant advantages on system reliability. The integrated reliability analysis is also explained using an event tree. The impact of the provision for forced circulation in the primary circuit on functional reliability is also studied with this procedure and it is found that the forced circulation capability helps to bring down the total decay heat removal failure probability by lowering the peak temperatures after the reactor shut down.     

非能动核电厂事故72 h后规程优化研究

事故72 h后规程是非能动核电厂特有的一类规程。为了评价非能动核电厂事故72 h后规程的必要性和充分性，并对其薄弱环节进行优化改进，本研究对示范项目非能动核电厂设计基准和运行规程体系的总体逻辑进行了分析，并基于事故72 h后规程的内容和结构，提出了评价事故72 h后规程必要性和充分性的通用方法，运用规程执行逻辑框图识别了规程执行过程中可能存在的风险项，并针对示范项目的规程提出了优化建议。分析结果表明，事故72 h后规程对于非能动核电厂的安全运行是充分必要的；其表达方式与规程体系结构相关；明确规程操作路径优先级、减少规程跳转能够提高规程执行效率。相关优化建议可为非能动核电厂事故后长期安全运行提供技术参考。      

中小型辐照装置的运行安全管理

分析了中小型^60Co辐照装置的易发事故，提出了运行控制的安全程序，并对运行程序的安全联锁提出具体要求，确保正常运行状态、故障状态和异常状态的辐射安全。最后，提出配套的安全防护管理制度。     

Potential failure mode and effects analysis for the ITER NB injector

The failure mode and effects analysis (FMEA) is a widely used analytical technique that helps in identifying and reducing the risks of failure in a system, component or process.The application of a systematic method like the FMEA was deemed necessary and adequate to support the design process of the ITER NBI (neutral beam injector). The approach adopted was to develop a FMEA at a general “system level”, focusing the study on the main functions of the system and ensuring that all the interfaces and interactions are covered among the various subsystems. The FMEA was extended to the whole NBI system taking into account the present design status. The FMEA procedure will be then applied to the detailed design phase at the component level, in particular to identify (or define) the ITER Class of Risk.Several important failure modes were evidenced, and estimates of subsystems and components reliability are now available. FMEA procedure resulted essential to identify and confirm the diagnostic systems required for protection and control, and the outcome of this analysis will represent the baseline document for the design of the NBI and NBTF integrated protection system.In the paper, rationale and background of the FMEA for ITER NBI are presented, methods employed are described and most interesting results are reported and discussed.     

Conceptual Requirements for Large Fusion Experiment Control, Data, Robotics, and Management Systems

The conceptual system requirements for the control, data, robotics, and project management (CDRM) system for the next generation of fusion experiments are developed by drawing on the success of the Tara control and data system. The requirements are described in terms of an integrated but separable matrix of well-defined interfaces among the various systems and subsystems. The study stresses modularity, performance, cost effectiveness, and exportability.     

Investigation on X-Radiation for 126 kV Vacuum Interrupters

When subjected to high voltages between opened contacts,vacuum interrupters may emit X-rays.In order to ensure that these are of an acceptable level,vacuum interrupters should comply with the limits for X-ray emission and the test procedures to be carried out to verify this based on relevant standards and specifications.In this paper,a comprehensive experimental study has been performed for 126 kV vacuum interrupters used in a transmission system to understand the X-radiation level and its influence by three main parameters,namely applied power-frequency voltage,contact gap and power-frequency voltage conditioning.The radiation instrument is an FJ347 radiometer and the X-radiation dose was measured at the power-frequency test voltage.These tests demonstrated that the X-radiation emission level for a 126 kV vacuum interrupter did not exceed the following:5 μSv per hour at a rated voltage of 126 kV and 150 μSv per hour at a power-frequency voltage of 230 kV at 1 m distance.The X-radiation dose increased with the applied power-frequency voltage increasing and decreased with the contact gap increasing.The Xradiation dose for 126 kV vacuum interrupters decreased by 57%after the conditioning procedure with a certain power-frequency voltage.During the conditioning procedure,the average value of the X-radiation dose was 4.49 mSv,which means if a professional conditions 180 interrupters per year,it will be safe at the 6.4 m distance.     

Considerations for in-service inspection on pressure vessel of 200MW nuclear heating reactor

The 200MW nuclear heating reactor adopts an integrated arrangement for the primary circuit. It is designed to be operated at lower temperature and lower pressure as compared to large reactors. A steel containment serves a barricade for the reactor pressure vessel. The pressure vessel has some safety characteristics, such as low stress level, low induced integral neutron flux, and high toughness etc. Among them, the most important is its LBB behavior. Based on the safety analysis for the pressure, the requirements and procedures of in-service inspection are layed-out accordingly.     

核电厂应急操作规程的现状及其发展趋势

本文叙述了核电厂应急操作规程的现状及不足之处,介绍了世界上在这方面的发展趋势:以基于征兆的应急操作规程来补充或替代基于事件的应急操作规程;采用现代微机数据处理技术改进规程的编制、修改、打印和显示;利用人工智能技术最终开发出核电厂运行专家系统,以帮助运行人员处理事故,减少人因错误。     

Computerised defect sizing using ALOK analysis in shrunk-on wheel of turbine rotor

A computerised defect evaluation system using an advanced ultrasonic technique for shrunk-on wheels of a turbine rotor has been developed. The flaw image was reconstructed by a modified ALOK method with attention to the tip diffraction echoes of stress corrosion cracking (SCC). The results of the application of this system to the mock-up wheel with artificial slits and real SCC cracks show that this system is very useful for crack sizing and locating.