Safety is a crucial aspect to deal with when designing a nuclear reactor. Among all of the possible accidental transients considered in the section Design-Basis Accidents (DBAs) of the Probabilistic Risk Assessments (PRAs), the in-vessel Loss of Coolant Accident (LOCA) is definitely one of the most critical, because of the significant damages caused to the components inside the Vacuum Vessel (VV). In this field proper system-level codes already exist. Due to their intrinsic lumped nature, these tools show good performances when studying the behavior of the entire system, but they are completely incapable of predicting localized quantities that are instead fundamental to assess the integrity of the VV in the most conservative way as possible.