## 1 Introduction  
The Lead-Cooled Fast Reactor (LFR) is one of the six nuclear reactor concepts selected by the Generation IV International Forum (GIF) to sustain the R&D of the next generation of nuclear energy systems [1]. The LFRs are cooled by Heavy Liquid Metals (HLM) such as lead or Lead-Bismuth Eutectic (LBE), employing a fast-neutron spectrum and a closed fuel cycle for efficient conversion of fertile uranium and management of actinides. The main advantages of using HLMs as coolant are [2]:  
* high thermal inertia;
* very low absorption and scattering cross sections;
* high boiling points allowing a low-pressure operation;good natural circulation due to the high density. Compared to sodium, lead and LBE has the key advantage of being an inert coolant without exhibiting violent chemical reactions with air and water. Therefore, by using lead or LBE the steam generator is located inside the reactor vessel, generating an integral compact system with the core and the coolant pump. As a shortcoming, lead and LBE are both weaker absorbers and moderators than sodium, allowing a larger amount of coolant in the core, forming wider cross sections of the flow paths and thus increasing the thermal inertia of the pool. Furthermore, the coolant is characterized by relatively low velocities which are necessary to mitigate the structural erosion. As a consequence, the pressure drops are reduced and the natural circulation is improved. Compared to lead, the advantage of LBE is the decreased freezing risk determined by lower melting temperature [1; 2; 3; 4]. Unfortunately, operational experience of LFRs is limited and further investigation is required to provide a larger amount of information regarding several aspects of the HLM systems. In particular, both the thermal-hydraulic behavior during normal operation and the scenarios generated by severe accident are key points for the designers. Besides the necessity of experimental campaigns, numerical analyses are strongly required to investigate the thermal-hydraulic aspects of HLMs. In this framework, the two projects thermal-hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors (ESSAME) and MYRRHA Research and Transmutation Endeavour (MYRTE) of HORIZON 2020, were launched. SESAME project was established to support the development of European liquid metal fast reactors, while the MYRTE project aims at studying the transmutation of high-level waste at industrial scale through the development of the Multi-Purpose Hybrid Research Reactor for High-Tech Applications (MYRRHA) research facility [3]. Inside these two projects, both experimental campaigns and numerical analyses were planned. Among the available numerical tools, the System Thermal Hydraulic (STH) codes are the most used because of their solid background. In the present work the STH code RELAP5/Mod3.3 [5; 6] was used to support the experimental campaign conducted at the ENEA (Energia Nucleare ed Energie Alternative) Brasimone Research Centre on the CIRColazione Eutettico- Heavy liquid mEtal pRessurized water cOuled tubes (CIRCE-HERO) facility [7; 8]. The version of RELAP5/Mod3.3 employed for this purpose was modified at Laboratory of Numerical Simulation for Nuclear Thermohydraulics of the University of Pisa by including the thermodynamic properties of lead and LBE, adding the updated transport properties for viscosity, thermal conductivity, surface tension and specific convective heat transfer correlations for liquid metals [9]. In particular, the thermodynamic properties of LBE were implemented according to the [PERSON]'s work [10], while heat transfer correlations developed for liquid metals by [PERSON] [11] and [PERSON] [12] were implemented as well. Lacking the physic and specific correlations for transported inondensable gases in liquid metals, the tables and flow regime maps developed for water and already implemented in the code were assumed as a good first approximation.