This paper describes an original solution of core catcher to managing the in vessel retention of the Corium in the accidental event of the core meltdown. The solution envisaged intends to verify the possibility of managing the accidental event within the pressure vessel, ensuring that the CORIUM is confined and cooled. The core catcher, elaborated at the DIMNP, is made of a ceramic pebbel bed (Alumina Al 2 O 3 ) contained in a metallic or Ceramic Matrix Composite (CMC) structure. The paper illustrates a theoretical model to simulate the thermal-mechanical behaviour of the pebble beds under extremely high loads, developed by the authors. This model has been used to design the core catcher and to determine the effective conductivity and the effective stiffness of the pebble bed. These values have been used in order to implement a numerical model of the core catcher. The results of the thermal and mechanical coupled simulation have permitted to determine the maximum time that the core catcher could resist and the mechanical resistance of the core catcher in the case of RPV external or internal cooling. The preliminary analyses performed have emphasised the good performance of pebble bed core catcher in order to mitigate the envisaged severe accident.

The intent of this paper is the presentation and discussion of a methodology for the evaluation and analysis of seismic loads effects on a nuclear power plant. To help in focussing the presented methodology, a preliminary simplified analysis of an integral, medium size next generation PWR reactor structure (IRIS project, an integral configuration PWR under study by an international group) was considered as an application example also for models/codes evaluation. The performed preliminary seismic analysis, even though by no means complete, is intended to evaluate the method of calculating the effects of dynamic loads propagation to the reactor internals for structural design as well as geometrical and functional optimisation purposes. To this goal, finite element method and separated (sub) structures approaches were employed for studying the overall dynamic behaviour of the nuclear reactor vessel. The analysis was set up by means of numerical models, implemented on the MARC FEM code, on the basis of Design Response Spectra as indicated on the relevant rules for Nuclear Power Plants (NRC 1.60) design. The seismic analysis is indented to evaluate the dynamic loads propagated from the ground through the Containment System and Vessel to the Steam Generator’s tubes.