Thermal Hydraulic Transient Analysis of Core Bypass Event in ALFRED

Generation Four (Gen-IV) reactors are a family of innovative and reliable nuclear technologies with high potential and long-term benefits in terms of sustainability. Lead Cooled Fast Reactors (LFRs) are one of the most promising systems in the Gen-IV group. The use of lead as heat transfer fluid with high boiling temperature (more than 1700 °C) at atmospheric pressures assures enhanced safety, higher thermal efficiencies (more than 40%) and lower costs. ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) is a 300 MWth LFR pool demonstrator produced by a European consortium led by ANSALDO Nucleare and used to evaluate the applicability of the European LFR knowledge and technology. In the context of the safety assessment of ALFRED, the Department of Civil and Industrial Engineering at the University of Pisa and ANSALDO started a system thermal-hydraulic (SYS-TH) investigation to simulate the accidental event of a leak/break in the internal structure of the reactor vessel. SYS-TH is a reduced order modelling strategy characterised by a coarse grained description of the main components in the system and the ability to produce reliable results with a relatively low computational cost. In this thesis, a SYS-TH RELAP5 nodalization of the ALFRED primary system has been setup and used to simulate the steady state of the reactor, at stage 2 (medium temperature). Then, leaks of the internal structure with different sizes were simulated to perform a sensitivity study on the break diameter. The results would provide useful information to guide the safety analysis of the reactor and diagnose the effects caused by these events. In a future perspective, these outcomes would be combined with more detailed methodologies and included in a multi-scale/multi-physics analysis.