Development of task analysis method for operator tasks in main control room of an advanced nuclear power plant

Task analysis methods provide an insight for quantitative and qualitative predictions of how people will use a proposed system, though the different versions have different emphases. Most of the methods can attest to the coverage of the functionality of a system and all provide estimates of task performance time. However, most of the tasks that operators deal with in a digital work environment in the main control room of an advanced nuclear power plant require high mental activity. Such mental tasks overlap and must be dealt with at the same time; most of them can be assumed to be highly parallel in nature. Therefore, the primary aim to be addressed in this paper was to develop a method that adopts CPM-GOMS (cognitive perceptual motor-goals operators methods selection rules) as the basic pattern of mental task analysis for the advanced main control room. A within-subjects experiment design was used to examine the validity of the modified CPM-GOMS. Thirty participants participated in two task types, which included high- and low-compatibility types. The results indicated that the performance was significantly higher on the high-compatibility task type than on the low-compatibility task type; that is, the modified CPM-GOMS could distinguish the difference between high- and low-compatibility mental tasks.     

Empirical investigation of communication characteristics under a computer-based procedure in an advanced control room

With the development of computer-based control rooms including a computer-based procedure (CBP), shift supervisors (SSs) can directly access plant information through a CBP or personal displays instead of depending on other board operators (BOs) to obtain plant information. In relation to this change, we examined the characteristics of SS inquiry patterns in the computer-based control rooms of nuclear power plants during emergency situations. Operator behaviors and speech patterns were observed and analyzed through several experiments on simulated accident scenarios in a full-scope simulator for an advanced computer-based main control room. We found that the SS inquiry in the advanced control room had less dependency on the BOs, and that the inquiry patterns varied according to the operators and operating dates. From these findings, the necessity to establish communication standards under computer-based control rooms was discussed with some recommendations. Another requirement to reduce the cognitive workload of SSs was also discussed.     

Empirical investigation of workloads of operators in advanced control rooms

This paper compares the workloads of operators in a computer-based control room of an advanced power reactor (APR 1400) nuclear power plant to investigate the effects from the changes in the interfaces in the control room. The cognitive–communicative–operative activity framework was employed to evaluate the workloads of the operator's roles during emergency operations. The related data were obtained by analyzing the tasks written in the procedures and observing the speech and behaviors of the reserved operators in a full-scope dynamic simulator for an APR 1400. The data were analyzed using an F-test and a Duncan test. It was found that the workloads of the shift supervisors (SSs) were larger than other operators and the operative activities of the SSs increased owing to the computer-based procedure. From these findings, methods to reduce the workloads of the SSs that arise from the computer-based procedure are discussed.     

AP1000型核电厂主控室内漏示踪气体试验方法简析

本文针对AP1000型核电厂主控室内漏示踪气体试验,在调查了美国相关标准及从事内漏示踪气体试验的公司的技术水平基础上,对浓度衰减法和恒量注入法在AP1000型核电厂应用时各自的优缺点进行了分析,并提出了相应的改进建议。     

核电厂主控室取风口监测仪表设置研究

核电厂主控室取风口监测仪表是为了监测取风口处的放射性核素浓度或环境剂量率,并与通风系统联动,进行取风口的切换。在事故后根据正常取风口处的剂量率监测仪表的监测数据可从正常通风系统切换至带过滤功能的应急通风系统,根据两个应急取风口处的β活度监测仪表的监测数据选择污染物浓度较低的应急取风口。因此取风口处监测仪表的设置对事故后保证主控室区域的可居留性至关重要。本文针对主控室通风系统正常取风口和应急取风口的监测仪表的设置进行研究,利用某厂址一整年气象数据,根据华龙一号机型典型事故源项、通风系统设计性能及主控室建筑结构特征对事故情况下从正常通风系统切换到应急通风系统的报警阈值进行分析计算,并结合监测仪表的响应特征,针对大气释放阀释放、安全壳释放和高架释放3种释放类型,分析了相应事故工况下主控室双取风口在不同的切换间隔条件下主控室工作人员所受剂量,最后给出推荐的主控室双取风口最短切换时间间隔。本文对取风口监测仪表设置方法的探索和研究为未来同类主控室通风系统监测的设计和优化提供了参考。     

事故工况下主控室应急新风系统碘吸附器性能试验研究

为了研究在一次事故期间主控室应急新风系统碘吸附器的效率变化情况,开展了浸渍活性炭的中毒老化试验、碘吸附器辐照试验和活性炭效率变化试验,通过综合各种试验因素,以制得的初始效率为99.90%、99.97%、99.98%三种炭样分别在温度70 ℃±1 ℃、相对湿度40%±2%和温度44 ℃±1 ℃、相对湿度95%±1.5%条件下开展了长时间的过滤效率变化试验。综合各项试验结果得出,浸渍活性炭存在自然老化现象,乙酸等物质会加速活性炭的中毒老化;辐照对浸渍活性炭的过滤效率无影响;在严重事故工况下,主控室应急新风系统碘吸附器的过滤效率在168 h 内没有下降。     

60Co集装箱CT检测系统主控站的设计与实现

简要介绍了^60Co集装箱CT检测系统的原理,并详细介绍了主控站软件的功能、编程技术和实现方法。该软件在实际运行中取得了很好的效果。     

核电厂土建施工三大原材料质量控制

核电厂的建造运行是一项系统工程。为了保证核电厂的质量安全,从设计立项到运行使用,每一个环节的工作都对质量安全产生不同程度影响。与之相关的每一步工作都必须做到质量受控。从核电厂建造过程中土建施工环节中的原材料质量控制入手,着重分析了土建施工中钢材、混凝土材料、模板材料三大主材如何实施质量控制和管理。     

An optimal range of information quantity on computer-based procedure interface design in the advanced main control room

The quantification of information in the interface design is a critical issue. Too much information on an interface can confuse a user while executing a task, and too little information may result in poor user performance. This study focused on the quantification of visible information on computer-based procedures (CBPs). Levels of information quantity and task complexity were considered in this experiment. Simulated CBPs were developed to consist of three levels: high (at least 10 events, i.e. 3.32 bits), medium (4–8 events, i.e. 2–3 bits), and low information quantity (1 or 2 events, i.e. 0 or 1 bits). Task complexity comprised two levels: complex tasks and simple tasks. The dependent variables include operation time, secondary task performance, and mental workload. Results suggested that medium information quantity of five to eight events has a remarkable advantage in supporting operator performance under both simple and complex tasks. This research not only suggested the appropriate range of information quantity on the CBP interface, but also complemented certain deficient results of previous CBP interface design studies. Additionally, based on results obtained by this study, the quantification of information on the CBP interface should be considered to ensure safe operation of nuclear power plants.     

CPR1000核岛主系统设备双机抬吊旋转法吊装工艺浅析

核岛主系统设备吊装是核电站核岛设备安装工作的重要环节。而重型设备吊装工艺方法复杂、技术难度大、安全技术高,需要耗用大量的人力物力。因此,核岛重型设备的吊装是核电站建设的重大关键技术问题。在核电蓬勃发展的今天,总结成熟的吊装技术,创造新的吊装工艺,推动核电重型设备吊装技术的发展,有着积极的现实意义。重型设备吊装技术作为一种社会财富,应该加以总结、提高和推广,以起到提供借鉴,开阔思路,指导吊装施工的作用。     

Development of hovering control system for an underwater vehicle to perform core internal inspections

In this paper, a hovering control system for an underwater vehicle is proposed to support core internal inspections. The system adopted a localization part and a thruster control part. The former utilizes a map-matching method, referring cross-sectional shape data cut from a three-dimensional computer aided design (CAD) and structural shapes measured by a laser range system for horizontal positioning. A pressure sensor provides vertical positioning. The latter utilizes the thrust vector control, or reference thrust vectors are converted to each propeller thrust based on the vehicle's geometric structure. Experiments to evaluate performance of the proposed system were implemented at a mock-up of the reactor bottom part. As a result, it was confirmed that the position was detected with an accuracy of 48 mm, and for a flow velocity of 200 mm/s, it was verified that the vehicle hovered within 77 mm of a target point. Therefore, core internal inspections can be stably carried out even where there is external force caused by water convection flow.     

基于Labwindows/CVI的氡室测控系统设计

针对目前氡室测控系统存在的问题，本文提出了基于CMOSense技术的SHT75数字式温湿度传感器和LaWindows／CVI虚拟仪器开发环境的设计思想。运用虚拟仪器本身提供的库函数、数据库技术和模块化的设计思想开发了氡室测控系统。该系统很好地完成了氡室里重要参数的实时全程控制，并且具有显示直观、反应迅速，性价比高等特点，因此具有一定的使用价值。     

Development and anti-swing control of an automated measurement robot system for multi-stud tensioning machine

During nuclear power plant maintenance, the multi-stud tensioning machine is used to perform opening/sealing the cap of the reactor pressure vessel. This process incorporates elongations of 58 studs, whose extension values are monitored in real time by measurement meters. Conventionally, the placements of the meters are performed by human labor, which is time consuming and radioactively hazardous. In this paper, we introduce an automated measurement robot system, consisting of 58 node robots and multiple field bus based distributed control devices, to complete meter placement and data acquisition tasks without human involvement in the hazardous working site. In order to eliminate the swing phenomenon of the wire-driven meter adaptor on the robot distal end, extra-insensitive input shaper is employed for robot motion control, thus saving the overall operation time from traditionally over 10 minutes to less than 22 s.     

Human reliability analysis for operators in the digital main control rooms of nuclear power plants

ABSTRACT

As the main control room of nuclear power plants (NPPs) has been gradually digitized, new human reliability problems may emerge because of a series of new changes in the cognitive processes, behavioral patterns, and error mechanisms of operators. Aiming to address this situation, this paper proposes a method as guidance for human reliability analysis (HRA) of different cognitive Stages. This method first constructs the influencing factors of three cognitive processes, including monitoring, decision-making, and execution of actions, and then evaluates the weights of these influencing factors through an analytic hierarchy process (AHP). In this study, the parameters used in the proposed HRA method were determined by analyzing the test data obtained from a simulation model, and the results demonstrated the rationality and feasibility of the proposed method. A case example using this HRA method was given in which the human error probabilities at three stages in a nuclear power plant (NPP) steam generator tube ruptures (SGTR) accident were obtained. In summary, the proposed method is a simple and feasible HRA tool that can be applied in digital NPP main control rooms (MCRs).     

Real-time measurement and feedback control of ion temperature profile and toroidal rotation using fast CXRS system in JT-60U

A fast charge exchange recombination spectroscopy (CXRS) system has been developed for the real-time measurement and feedback control of ion temperature (T_i) profile and toroidal rotation velocity (V_t) in JT-60U. In order to control T_i and V_t in real-time, the charge exchange recombination spectroscopy with high time resolution, the real-time processor system, and the real-time control system have been developed. Utilizing this system, real-time control of the T_i gradient between r/a ∼ 0.25-0.5 has been demonstrated with neutral beams at high beta plasmas (normalized beta β_N ∼ 1.6-2.8). The strength of the internal transport barrier is controlled. Moreover, the real-time control of V_t has been demonstrated from counter (anti-parallel to the plasma current, I_p) to co (parallel to the I_p) direction. Then the behavior of edge localized mode (ELM) is changed by controlling the V_t.     

核电站核岛物项质量控制信息化实现

根据核电站核岛物项质量管理模式,阐述了核电站核岛物项质量控制信息化实现的设计思路、实现方法和一些关键技术,说明了核岛物项质量控制方式可以提高核岛质量、保证核电站的安全、降低核电站建造和运营成本。     

An estimation of an operator's action time by using the MARS code in a small break LOCA without a HPSI for a PWR

To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the ‘ASME PRA standard’ also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS code has been developed for the PSA of the Korea standard nuclear power plant (KSNP). This study has proposed an estimation method for an operator's action time by using the MARS platform. The typical example case is a small break loss of coolant accident without the high pressure safety injection system, which is one of the most important accident sequences in the PSA of the KSNP. Under the given accident sequence, the operator has to perform a recovery action known as a fast cooldown operation. This study focuses on two aspects regarding an operator's action; one is how they can operate it under some restrictions; the other is how much time is available to mitigate this accident sequence. To assess these aspects, this study considered: (1) the operator's action model and (2) the starting time of the operation. To show an effect due to an operator's action, three kinds of control models (the best-fitting, the conservative, and the proportional-integral) have been assessed. This study shows that the developed method and the platform are useful tools for this type of problem and they can provide a valuable insight related to an operator's actions.     

Developing architecture for upgrading I&C systems of an operating nuclear power plant using a quality attribute-driven design method

This paper presents the architecture for upgrading the instrumentation and control (I&C) systems of a Korean standard nuclear power plant (KSNP) as an operating nuclear power plant. This paper uses the analysis results of KSNP's I&C systems performed in a previous study. This paper proposes a Preparation–Decision–Design–Assessment (PDDA) process that focuses on quality oriented development, as a cyclical process to develop the architecture. The PDDA was motivated from the practice of architecture-based development used in software engineering fields. In the preparation step of the PDDA, the architecture of digital-based I&C systems was setup for an architectural goal. Single failure criterion and determinism were setup for architectural drivers. In the decision step, defense-in-depth, diversity, redundancy, and independence were determined as architectural tactics to satisfy the single failure criterion, and sequential execution was determined as a tactic to satisfy the determinism. After determining the tactics, the primitive digital-based I&C architecture was determined. In the design step, 17 systems were selected from the KSNP's I&C systems for the upgrade and functionally grouped based on the primitive architecture. The overall architecture was developed to show the deployment of the systems. The detailed architecture of the safety systems was developed by applying a 2-out-of-3 voting logic, and the detailed architecture of the non-safety systems was developed by hot-standby redundancy. While developing the detailed architecture, three ways of signal transmission were determined with proper rationales: hardwire, datalink, and network. In the assessment step, the required network performance, considering the worst-case of data transmission was calculated: the datalink was required by 120 kbps, the safety network by 5 Mbps, and the non-safety network by 60 Mbps. The architecture covered 17 systems out of 22 KSNP's I&C systems. The architecture is implementable with the equipment developed in South Korea. The architecture can be used as a model to upgrade the existing I&C systems in a planned, large-scale, and one-shot manner. A more detailed architecture down to software level will be developed in the future.     

An overview of instrumentation and control systems of a Korea standard nuclear power plant: A signal interface standpoint

This paper presents an overview of instrumentation and control (I&C) systems of a pressurized water reactor (PWR) type nuclear power plant (NPP) in Korea. Yonggwang unit 3, which was constructed as a basis model for a Korea standard nuclear power plant (KSNP), is selected as an example for the presentation. This overview is derived from analyzing the I&C systems based on a top-down approach. The I&C systems consist of 30 systems. The 183 I&C cabinets are also analyzed and mapped to the systems. The overview is focused on an interface between the systems and the cabinets. This information will be used to understand the implementation of the I&C systems and to group the systems for an upgrade.     

Axial offset control of PWR nuclear reactor core using intelligent techniques

Improved load following capability is one of the main technical performances of advanced PWR (APWR). Controlling the nuclear reactor core during load following operation encounters some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking, while the core is subject to large and sharp variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent of core power peaking, in form of a practical parameter. This paper, proposes a new intelligent approach to AO control of PWR nuclear reactors core during load following operation. This method uses a neural network model of the core to predict the dynamic behavior of the core and a fuzzy critic based on the operator knowledge and experience for the purpose of decision-making during load following operations. Simulation results show that this method can use optimum control rod groups maneuver with variable overlapping and may improve the reactor load following capability.