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Integrated software safety analysis method for digital I&C systems
Authors:Hui-Wen Huang  Chunkuan Shih  Swu Yih  Ming-Huei Chen
Affiliation:1. Institute of Nuclear Energy Research, No. 1000, Wenhua Road, Chiaan Village, Longtan Township, Taoyuan County 32546, Taiwan;2. Department of Engineering and System Science, National Tsing-Hua University, 101, Section 2 Kuang Fu Road, Hsinchu, Taiwan;3. Department of Computer Science and Information Engineering, Ching Yun University, 229, Chien-Hsin Road, Jung-Li City, Taiwan
Abstract: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.
Keywords:ABWR  advanced boiling water reactor  ADS  automatic depressurization system  BTP  branch technical position  CCF  common-cause failure  CPU  central processing unit  D3  diversity and defense-in-depth  ECCS  emergency core cooling system  EPROM  erasable programmable read only memory  ESFAS  engineered safety features actuation system  FMEA  failure modes and effects analysis  FTA  fault tree analysis  HPCF  high pressure core flooder  I&  C  instrumentation and control  IBM  international business machines corporation  INER  institute of nuclear energy research  LOCA  loss of coolant accident  LPFL  low pressure core flooder  NPP  nuclear power plant  NRC  nuclear regulatory commission  MCR  main control room  NTHU  National Tsing Hua University  PCTran  personal computer transient analyzer  PHA  preliminary hazard analysis  PRA  probabilistic risk assessment  rcic  reactor core isolation cooling  RFC  recirculation flow control system  RHR  residual heat removal system  RPS  reactor protection system  RPV  reactor pressure vessel  RSP  remote shutdown panel  Rx  Reactor  SAR  safety analysis report  SCM  software configuration management  SSA  software safety analysis  SVV  software verification and validation  TAF  top of active fuel  UML  unified modeling language  VDU  video display unit
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