Creep rupture assessment for Level-2 PSA of a 2-loop PWR： accounting for phenomenological uncertainties

The Level-2 probabilistic safety assessment （PSA） of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system compo- nents during the course of high pressure severe accident sequences. The present paper covers this technical issue and tries to quantify its associated phenomenological uncertainties for the development of Level-2 PSA. A framework is proposed for the formal quantification of uncertainties in the Level-2 PSA model of a PWR type nuclear power plant using an integrated deterministic and PSA approach. This is demonstrated for estimation of creep rupture failure probability in station blackout severe acci- dent of a 2-1oop PWR, which is the representative case for high pressure sequences. MELCOR 1.8.6 code is employed here as the deterministic tool for the assessment of physical phenomena in the course of accident. In addition, a MATLAB code is developed for quantification of the probabilistic part by treating the uncertainties through separation of aleatory and epistemic sources of uncertainty. The probability for steam generator tube creep rupture is estimated at 0.17.