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Digital instrumentation and control failure events derivation and analysis for advanced boiling water reactor

机译:先进沸水反应堆数字仪表和控制故障事件的推导与分析

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[[abstract]]This research adopted Personal Computer Transient Analyzer- Advanced Boiling Water Reactor version (PCTran-ABWR) simulation computer code to analyze the software safety issue for a generic ABWR. A number of postulated instrumentation and control (I and C) system software failure events were derived to perform the dynamic analyses. The basis of event derivation includes the published classification for software anomalies, the digital I and C design data of ABWR, chapter 15 accident analysis of generic safety analysis report (SAR), and the reported nuclear power plant I and C software failure events. For the purpose of enhancing the ABWR major control systems simulation capability, this research incorporated MATLAB into PCTran-ABWR to improve the pressure control system, feedwater control system, recirculation control system, and automated power regulation control system. As a result, the software failure of these digital control systems can be properly simulated and analyzed. Moreover, via an internal tuning technique, the modified PCTran-ABWR can precisely reflect the characteristics of the power-core flow map. Hence, in addition to transient plots, the analysis results can then be demonstrated on the Power-Core Flow Map. The case study of this research includes (I) the software common mode failures analysis for the major digital control systems; and (2) postulated ABWR digital I and C software failure events derivation from the actual happening of non-ABWR digital I and C software failure events, which were reported to Licensee Event Report (LER) of US Nuclear Regulatory Commission (USNRC) or Incident Reporting System (1RS) of International Atomic Energy Agency (IAEA). These events were analyzed by PCTran-ABWR. Conflicts among plant status, computer status, and human cognitive status are successfully identified. The operator might not easily recognize the abnormal condition, because the computer status seems to progress normally. However, a well trained operator can become aware of the abnormal condition with the inconsistent physical parameters; and then can take early corrective actions to avoid the system hazard. This paper also discusses the advantage of Simulation-based method, which can investigate more in-depth dynamic behavior of digital I and C system than other approaches. Some unanticipated interactions can be observed by this method. © 2006 IEEE.
机译:[[摘要]]本研究采用了个人计算机瞬态分析仪-高级沸水反应堆版本(PCTran-ABWR)模拟计算机代码来分析通用ABWR的软件安全性问题。推导了许多假定的仪器和控制(I和C)系统软件故障事件,以执行动态分析。事件推导的基础包括已发布的软件异常分类,ABWR的数字I和C设计数据,通用安全分析报告(SAR)的第15章事故分析以及已报告的核电厂I和C软件故障事件。为了增强ABWR主控制系统的仿真能力,本研究将MATLAB合并到PCTran-ABWR中,以改进压力控制系统,给水控制系统,再循环控制系统和自动功率调节控制系统。结果,可以适当地模拟和分析这些数字控制系统的软件故障。而且,通过内部调整技术,改进后的PCTran-ABWR可以精确反映功率核流图的特征。因此,除了瞬变图外,分析结果还可以在Power-Core Flow Map上得到展示。本研究的案例研究包括:(I)主要数字控制系统的软件共模故障分析; (2)假定ABWR数字I和C软件故障事件是由非ABWR数字I和C软件故障事件的实际发生而引起的,并已报告给美国核监管委员会(USNRC)或事件的被许可方事件报告(LER)国际原子能机构(IAEA)的报告系统(1RS)。 PCTran-ABWR对这些事件进行了分析。已成功识别出工厂状态,计算机状态和人类认知状态之间的冲突。操作员可能不容易识别异常情况,因为计算机状态似乎正常进行。但是,训练有素的操作员可能会意识到物理参数不一致的异常情况。然后可以及早采取纠正措施以避免系统危害。本文还讨论了基于仿真的方法的优点,该方法比其他方法可以更深入地研究数字I和C系统的动态行为。通过这种方法可以观察到一些意外的相互作用。 ©2006 IEEE。

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    HuangHui-Wen;

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  • 年度 2010
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