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Development and diversity and defense-in-depth application of ABWR feedwater pump and controller model
Authors:Hui-Wen Huang [Author Vitae]  Chunkuan Shih  Ming-Huei Chen
Affiliation:a Institute of Nuclear Energy Research, No. 1000, Wenhua Road, Chiaan Village, Longtan Township, Taoyuan County, 32546, Taiwan, ROC
b Institute of Nuclear Engineering and Science, National Tsing-Hua University, 101, Section 2 Kuang Fu Road, Hsinchu, Taiwan, ROC
Abstract:This work developed an advanced boiling water reactor (ABWR) feedwater pump and controller model, which was incorporated into Personal Computer Transient Analyzer (PCTran)-ABWR, a nuclear power plant simulation code. The feedwater pump model includes three turbine-driven feedwater pumps and one motor-driven feedwater pump. The feedwater controller includes a one-element/three-element water level controller and a specific feedwater speed controller for each feedwater pump. The performance tests, including step change of dome pressure, feedwater pumps transfer, inadvertent closure of all turbine control valves, and one feedwater pump trip at 100% power, demonstrate the feasibility of dynamic response of stand-alone model and incorporated model. Furthermore, a diversity and defense-in-depth analysis is performed to demonstrate the feasibility for motor-driven feedwater pump as an emergency core cooling system (ECCS) automatic diverse back-up. In Lungmen nuclear power plant (NPP), a diverse manual initiation means for the high pressure core flooder (HPCF) loop C is designed as the back-up of digitalized engineered safety features actuation system (ESFAS). If the motor-driven feedwater pump (MDFWP) can be an automatic digital diverse back-up for ESFAS, Lungmen NPP would be more robust to defend against software common-cause failure (CCF).
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  DNBR  departure from nuclear boiling ratio  ECCS  emergency core cooling system  ESFAS  engineered safety features actuation system  FMEA  failure modes and effects analysis  FTA  fault tree analysis  FWP  feedwater pump  HPCF  high pressure core flooder  I&  C  instrumentation and control  INER  Institute of Nuclear Energy Research  LOCA  loss of coolant accident  LPFL  low pressure core flooder  MD  motor-driven  NPP  nuclear power plant  NRC  Nuclear Regulatory Commission  NTHU  National Tsing Hua University  PCTran  Personal Computer Transient Analyzer  PHA  preliminary hazard analysis  PRA  probabilistic risk assessment  RCIC  reactor core isolation cooling  RHR  residual heat removal system  RPS  reactor protection system  RPV  reactor pressure vessel  SCM  software configuration management  SSA  software safety analysis  SV&  V  software verification and validation
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