• preliminary safety analysis
• NRC
• RELAP5
• IRIS

The present master degree thesis in Nuclear Engineering deals with the preliminary safety assessment of the IRIS nuclear reactor using the Relap5/Mod 3.3 thermal-hydraulic code. Most of the present work has been performed during a six months stay period at Westinghouse Science and Technology Center (STC) in Pittsburgh (PA).
After an overview of the current nuclear power plant situation in the United States following TMI and Chernobyl accidents, the result of the U.S.DOE efforts to plan and deploy a safe NSSS in the Generation IV proposal are presented, with attention focused on the Westinghouse IRIS project. This project is synthetically described, with the methodology developed for system analysis, and the technological and plant design innovations that are in conformity with the revolutionary requirements of Generation IV nuclear power plants.
A summary of the RELAP5 nodalization and a detailed description of the procedures and assumptions used in the definition of the main components of the IRIS system at March 2003 are then presented. The attention is focused on the contributes that has been provided in this thesis: the IRIS integral reactor coolant system residence time calculation, and the development of a methodology to generate IRIS initial steady state conditions to assist the safety assessment procedure.
The previous section provides the basis to the preliminary study of some of the main design basis accidents of IRIS reactor presented in the final part of this work, and performed following the guidelines of the NRC Standard Review Plan and AP600 safety analysis report.
Detailed analysis of the most significative accidents of decrease in heat removal by the secondary system and decrease in reactor coolant system flow has been developed with RELAP5 computer code model, and the are shown in the following:


· Decrease in heat removal by the secondary system
o Loss of external load and turbine trip
o Loss of offsite power and loss of normal feedwater
o Feedline Break
· Decrease in reactor coolant system flow
o Partial loss of flow
o Complete loss of flow
o Locked rotor