Fragility comparison analysis of CPR1000 PWR containment subjected to internal pressure

Evaluating the reliability of the containment structure of a nuclear power plant under an accident internal pressure is of importance in Design Basis Accident and Design Extension Condition analyses. To study the impact of the diverse requirements concerning material properties in various Nuclear Power Plant (NPP) design codes on the reliability of the CPR1000 PWR containment structure, its detailed full-sized three-dimensional Finite Element (FE) model with nonlinear material constitutions, accurate geometric features, and a complex reinforcement system is constructed, and a set of material property data is obtained through practical measurements at a certain CPR1000 power plant for comparison. Based on the Monte Carlo method, random properties of the main material of the containment are applied to produce 100 samples using two sets of NPP design codes (Chinese and European) and one set of practically measured data. Various damage patterns of the samples generated owing to the uncertainty in the material properties are distinguishable. Generally, a concrete crack near an equipment hatch causes containment functional failure and when pre-stressed tendons begin to yield, the containment structure would not continue to maintain its integrity. Based on the functional and structural failure criteria, the corresponding fragility curves are obtained, and the results suggest that irrespective of the differences in the definition of material properties in various NPP design codes, the calculated reliability is practically the same. The difference of internal pressures corresponding to 5% failure probability is less than 2.6%. However, the containment based on practically measured data shows a slightly higher and slightly lower reliability than that based on Chinese and European design codes, respectively.