Analysis of a sodium-cooled fast reactor during ULOF and UTOP transients using the AZTHLIM code

The main goal of this work is to perform an analysis of a Sodium-cooled Fast Reactor using the AZTHLIM code, simulating two transient events: an Unprotected Loss-of-Flow (ULOF) and un Unprotected Transient-over-Power (UTOP). The AZTHLIM code is a thermo fluid code whose main objective is to describe the processes of heat transfer between the fuel rod and the coolant, developed as part of the AZTLAN Platform project. The core design used in this work has the characteristics of European Sodium Fast Reactor (ESFR), the nominal power of 3600 M Wth and fuel based on mixed oxides of uranium and plutonium. For the transient analysis, it is applicate a reduction of the coolant flow to simulate a ULOF transient, using 50% and 10% of the nominal flow. In the case of UTOP transient, it is considered an increase of 100 pcm and 175 pcm in reactivity. For ULOF transients, the maximum sodium temperature obtained is 964.18 K, when the inlet flow is 10% of nominal flow. And for the UTOP transient, 875.72 K was the maximum temperature with an increase of 175 pcm. (Author)