A Study of Seismic Probability Safety Assessment for a Typical Three-Loop Pressurized Water Reactor Nuclear Power Plant

An introduction of the background, purpose and the methodology of seismic Probability Safety Assessment (PSA) for the typical three-loop pressurized water reactor nuclear power plant are firstly given. Seismic hazard curves for the site of the plant are developed through Probabilistic Seismic Hazard Analysis (PSHA). With the input of site specific Uniform Hazard Response Spectrums (UHRSs), Soil-Structure Interaction analysis is completed using multiple time-history analysis method. After the development of the seismic equipment list, a screening process is completed with screening criteria determined on the basis of seismic hazard curves and a rough estimate of the seismic risk level. For the structures and components not screened out, seismic fragility parameters are calculated following the separation of variables approach. A seismic PSA model is developed afterwards, in which a seismic pre-tree and several seismic event trees as well as relating fault trees are built to model the accident sequence in seismic events. The quantification of the model is done on the software of RiskSpectrum and its additional code for hazard analysis, RiskSpectrum.HazardLite. According to the results, seismic induced Core Damage Frequency (CDF) of the plant is 2.69×10⁻⁶/reactor year, and the range of Peak Ground Acceleration (PGA) of seismic events from 0.30g to 0.45g has the most significant risk contribution. Seismic induced loss of offsite power, loss of emergency AC power and DC power are important accidents for the plant. Electrical components, auxiliary feeding water tank, diesel generators and ceiling of main control room of the plant are important contributors to seismic risk.