The digitalized Instrumentation and Control (I&C) system of Nuclear power plants can provide more powerful overall operation capability, and user friendly man-machine interface. The operator can obtain more information through digital I&C system. However, while I&C system being digitalized, three issues are encountered: 1) software common-cause failure, 2) the interaction failure between operator and digital instrumentation and control system interface, and 3) the non-detectability of software failure. These failures might defeat defense echelons, and make the Diversity and Defense-in-Depth (D3) analysis be more difficult. This work developed an integrated methodology to evaluate nuclear power plant safety effect by interactions between operator and digital I&C system, and then propose improvement recommendations. This integrated methodology includes component-level software fault tree, system-level sequence-tree method and nuclear power plant computer simulation analysis. Software fault tree can clarify the software failure structure in digital I&C systems. Sequence-tree method can identify the interaction process and relationship among operator and I&C systems in each D3 echelon in a design basis event. Nuclear power plant computer simulation analysis method can further analyze the available backup facilities and allowable manual action duration for the operator when the digital I&C fail to function. Applying this methodology to evaluate the performance of digital nuclear power plant D3 design, could promote the nuclear power plant operation safety. The operator can then trust the nuclear power plant than before, when operating the highly automatic digital I&C facilities.
This research adopted Personal Computer Transient Analyzer-Advanced Boiling WaterReactor version (PCTran-ABWR) simulation computer code to analyze the software safety issue for a generic ABWR. A number of postulated instrumentation and control (I&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&C design data of ABWR, chapter 15 accident analysis of generic safety analysis report (SAR), and the reported nuclear power plant I&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 (1) the software common mode failures analysis for the major digital control systems; and (2) postulated ABWR digital I&C software failure events derivation from the actual happening of non-ABWR digital I&C software failure events, which were reported to Licensee Event Report (LER) of US Nuclear Regulatory Commission (USNRC) or Incident Reporting System (IRS) 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&C system than other approaches. Some unanticipated interactions can be observed by this method.
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