Fragility analysis of nuclear steam supply system components of a pressurized water reactor

The present paper deals with the fragility analysis of the nuclear steam supply system (NSSS) at Gosgen nuclear power plant (KKG), a three-loop pressurized water reactor (PWR). The analysis covers the stability and integrity of the main NSSS components and piping: reactor pressure vessel (RPV), steam generators (SG), reactor coolant pumps (RCP), pressurizer, main coolant lines, surge line, main feedwater and main steam lines (inside the reactor building). Furthermore, the fragility of the reactor trip is analyzed with respect to failure modes associated with the RPV internals (RPVI). The fragilities are based on a probabilistic soil-structure interaction analysis with a coupled model of the reactor building and the above mentioned NSSS components/piping, see Rangelow et al. (2019. Probabilistic SSI analysis of a reactor building to facilitate fragility assessment. Transactions, SMiRT-25, Charlotte, NC, USA). The ground motion is represented by time histories corresponding to the median uniform hazard spectra (UHS) with an annual probability of exceedance of 10(-4). The fragility analysis is performed according to the separation-of-variables method, following guidance documents such as EPRI (2009. Seismic fragility applications guide update, Report 1019200. Palo Alto, CA) and EPRI (2018. Seismic fragility and seismic margin guidance for seismic probabilistic risk assessments, Report 3002012994. Palo Alto, CA). For the RPV internals, the fragility is based on two non-linear analysis models: a refined one for the evaluation of the fuel assembly spacer grids and a more global one for the remaining failure modes. A combined fragility has also been derived for the RPV internals, as the union of the individual failure modes. The paper discusses the method, the input data and selected results of the analysis.