Analysis of the Thermal-Hydraulics of several Loss of Coolant Accidents in a generic Pressurized Water Reactor

Abstract

The need to ensure the safety in nuclear power plants for operators, public and environment make essential to take special care in design in order to avoid any hazard, improve structural reliability or minimize the damages in case of accident. The purpose of the present thesis is to analyse the thermal-hydraulics behaviour in a pressurized water reactor during a small break loss-of-coolant accident (SBLOCA), so as to implement different modelling options and parameters in ATHLET code. Several scenarios are considered based on different break sizes, leak orientations and hypotheses on the availability of emergency core cooling systems. In this model, the PWR consists of four independent coolant loops connected to the reactor on the primary side, and four independent feedwater loops on the secondary side to supply the steam generators. Furthermore, the core is divided into seven regions: three at the centre and four around the periphery. The results obtained in the validation of the steady state and the transient calculation show the physical phenomena expected in this kind of accidents, which reveals that the ATHLET code provides a good confidence basis for the study of SBLOCAs in a PWR.