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• ITER TBM Program plans to test 4 DEMO-relevant tritium breeding blanket systems. • There are two water-cooled and two helium-cooled Test Blanket Systems (TBS). • Each TBS is formed by various sub-systems with different functions and technologies. • The paper describes the on-going associated R&D and the main testing objectives. The ITER TBM Program foresees that two ITER equatorial ports are dedicated to the simultaneous operation and testing of mock-ups of four different concepts of tritium breeding blankets (TBB) ensuring tritium breeding self-sufficiency and extraction of high-grade heat for a demonstration fusion reactor (DEMO). The paper starts with the description of the four Test Blanket Systems (TBSs) for the initial configuration and the main on-going R&D carried out by the ITER Members (IMs) in support of the TBS designs. The second part addresses the main objectives of the ITER TBM program to be obtained during the testing campaigns in ITER in order to demonstrate the feasibility of the corresponding DEMO TBBs. It also addresses the main technological data/information that are expected to be obtained during the TBSs design and manufacturing phases and relevant for the DEMO breeding blanket designs. [ABSTRACT FROM AUTHOR]

• STEAM will experimentally validate the DEMO WCLL BoP water coolant systems. • STEAM is an experimental facility equipped with air coolers operating as heat sink. • The steady-state characterization of STEAM was performed by RELAP5/Mod3.3. • Critical aspects of the air coolers operation were investigated and discussed. • Preliminary control systems were tested for some of the main STEAM components. DEMO power station aims to demonstrate the generation of hundred MWs of electrical power from fusion reactions, then transmitted from the Tokamak reactor to the grid through the Balance of Plant (BoP). The design approach for the Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) leverages nuclear industry expertise but faces challenges due to DEMO pulsed operation and low-load periods. To assess the feasibility of these components, ENEA Experimental Engineering Division at Brasimone R.C. is designing STEAM, a facility investigating water technologies applied to the DEMO BB and BoP systems and components. STEAM is mainly composed of a primary system reproducing the DEMO WCLL BB PHTS thermodynamic conditions (15.5 MPa, 328–295 °C) and a secondary two-phase (liquid/steam) loop reproducing the DEMO power conversion system conditions (6.4 MPa, 238–300 °C). Experimental validation will reproduce steady-state and transient operation under DEMO-relevant conditions, including dedicated tests on the DEMO once through steam generator mock-up. STEAM objectives and description are presented in this paper, together with the RELAP5/Mod3.3 nodalization of the facility. The latter is used to, thermal-hydraulically characterize the facility behaviour. The outcomes of the steady-state qualification supported the optimization of the system layout appraising the performances of key components under the specified operating conditions. [ABSTRACT FROM AUTHOR]

• A lot of experience has been gained at CEA in designing neutron sensors to be applied to DEMO breeding blankets. • Self-powered neutron detectors, fission chambers, optical fission chamber, dosimetry with activation foils are considered. • An effort joining theoretical/simulation developments to experimental campaigns is needed for DEMO application. Neutron measurements are routinely performed in fission reactors. A lot of experience has been gained at CEA in designing, prototyping and testing neutron sensors and associated acquisition systems in several nuclear facilities with a wide range of demanding conditions. In this paper, the applicability of this experience to neutron diagnostics in the breeding blankets of DEMO is addressed, as a complement to the already planned diagnostic strategy. The breeding blankets are a key component of a fusion reactor such as DEMO. The expected performances and operational constraints are demanding, nevertheless some of those, as the neutron flux range and high temperature, are close to what is encountered in sodium-cooled fast neutron reactors. The strengths and drawbacks of neutron sensors are presented with respect to the performances and constraints. Self-powered neutron detectors and fission chambers benefit from a vast return of experiment in fission reactors and received recent developments in the framework of studies for sodium-cooled fast neutron reactors. The innovative optical fission chamber is discussed for its electromagnetic immunity and high dynamic range demonstrated up to seven decades. Dosimetry with activation foils is also considered for its ability to provide an absolute calibration and spectral unfolding. [ABSTRACT FROM AUTHOR]

The Broader Approach (BA) activities aim to complement the ITER project and to promote the early realization of fusion energy through research, development, and tests of technologies supporting the future demonstration fusion reactor (DEMO). These activities are implemented under the BA agreement, which was signed and ratified in 2007 between Euratom and the Government of Japan. In essence, the BA activities consist of three projects: a) The Satellite Tokamak Programme Project JT-60SA, the world's largest superconducting tokamak until ITER starts, aims to support the assembly, commissioning and preliminary operation of ITER, and carry out demonstration and optimisation of steady-state operation of advanced plasma configurations for DEMO; b) The International Fusion Energy Research Centre (IFERC), which comprises three different sub-projects: the DEMO Design Research and Development Coordination Centre to coordinate design and R&D on materials and components for DEMO, the Fusion Computer Simulation Centre for the simulations of fusion plasmas, the analysis of experimental data, modelling of ITER operation, and contribution to the design of DEMO, and the Remote Experimentation Centre to allow scientists to participate remotely in fusion experiments from its control room in Japan; c) The Engineering Validation and Engineering Design Activities for the International Fusion Materials Irradiation Facility (IFMIF/EVEDA) focused on the design and validation of key components needed for the future Fusion Neutron Source facilities to characterize materials envisioned for DEMO. After 13 years of fruitful collaboration, Euratom and Japan launched the second phase of the BA activities in 2020. It is focused on exploiting and enhancing the facilities that have already been built and on working more closely than ever with ITER as the latter advances towards the first plasma. This paper will present a synthesis of the achievements already obtained thanks to the three BA projects and will address the future developments planned in 2023 and beyond. [ABSTRACT FROM AUTHOR]

• STEAM will host the mock-up of the EU-DEMO Steam Generators. • STEAM primary loop has been investigated with RELAP5/Mod3.3 system code. • DEMO normal operations have been studied including pulse-dwell transitions. • Different primary loop management strategies have been numerically simulated. • Setting the primary average temperature minimizes the loop operative fluctuations. The Water Cooled Lithium Lead Breeding Blanket (WCLL BB) is a key candidate for the driver blanket of the European DEMO reactor, progressing toward its Conceptual phase by the end of 2027. To assess different water and lithium-lead technologies for the WCLL BB and Balance of Plant (BoP) systems, the Water-thermal-HYDRAulic (W-HYDRA) experimental platform is under development at the ENEA Brasimone Research Centre. Among the facilities constituting the new W-HYDRA multipurpose infrastructure, STEAM is going to experimentally investigate the DEMO WCLL BoP thermal-hydraulics, focusing on the Steam Generator (SG) of the Primary Heat Transfer Systems (PHTS), to qualify its performances and suitability under its unconventional operation. The paper aims at supporting the thermal-hydraulic characterization of the Steam Generator mock-up during the sudden power variations typical of a pulsed fusion reactor. The analyzed selected scenario is the operational transient dwell-pulse-dwell, which determines high thermal cycling and correspondent high thermo-mechanical stresses on the primary side components. Two control logics with their relative drawbacks have been analyzed with a RELAP5/Mod3.3 1-D model, the first regulating the primary side average temperature, the second monitoring the minimum one. The comparison of the two systems highlighted that neither approach leads to hazardous conditions for the facility. However, while the average temperature controller is characterized by reduced thermal stresses on the components, the minimum temperature controller is characterized by higher thermal gradients for the primary loop. Both methodologies will be tested in the dedicated experimental campaign, aiming at yielding insights and evaluations concerning control strategies applicable to the DEMO reactor. [ABSTRACT FROM AUTHOR]

The M1 mirrors are part of the in-vessel quasi-optical Electron Cyclotron beam propagation system of the DEMO nuclear fusion reactor. These four mirrors are all non-movable and with a plane reflecting surface. Based on their functionality they are divided into two groups, the Mid-Steering Antenna (MSA) and the Multi-Beam Mirror (MBM) antenna. The MSA is responsible for controlling the neoclassical tearing modes, while the MBM is available for plasma bulk heating, current drive and thermal instability control. These mirrors are designed to withstand the thermal loads generated during operation: neutron heating, ohmic losses and stray radiation caused by the mm-wave beam reflection. To fulfil the design requirements, efficient heat removal strategies based on water cooling principles minimizing pressure drop are implemented. This paper presents the pre-conceptual design of the MBM and MSA M1 mirrors, validating the structural and geometric boundary conditions, as well as the appropriate selection of materials. Using FEA tools, thermal-hydraulic and thermo-mechanical simulations were performed with the available thermal loads to optimize the cooling parameters and perform a first stress assessment on the two M1 mirrors. • Thermal-hydraulics and thermo-mechanical simulations performed on the M1 mirrors. • The cooling channels design is mainly based on parallel spirals at the beam spots. • Ohmic losses (170 GHz) and neutron heat load were considered as thermal loads. • The inlet conditions of the water cooling line are 75 °C and 3.5 MPa. • The design ensures acceptable temperature, pressure and deformation. [ABSTRACT FROM AUTHOR]

The International Fusion Materials Irradiation Facility-DEMO Oriented NEutron Source (IFMIF-DONES) is an accelerator-based neutron source that generates high-energy neutrons via stripping processes by focusing a high-energy deuteron beam on a fast-flowing liquid lithium jet. The neutrons produced are used for the irradiation of materials foreseen in DEMO, thus providing relevant data for the design and licensing of the future fusion reactor. The complexity of such a plant is managed by a central control system that guarantees the safe supervision and control of all operations. This paper summarizes the most recent developments in the design of the IFMIF-DONES plant's Central Instrumentation and Control Systems (CICS) after the completion of the preliminary design phase. In particular, the architecture of two of the main CICS systems (namely, MPS and SCS) is described in detail, focussing on the specific design choices recently proposed. For each system, the current status of design is presented, as well as the existing and future plans for their integration in a unique control framework. [ABSTRACT FROM AUTHOR]

• Review paper summarizing years of research on fusion materials. • Advanced modifications to F82H and EUROFER. • Summarised process to validate RAFM steels for operation in DEMO IVCs. • Strategic direction for EU-DEMO materials data gathering presented. This work addresses an overview of the recent progress, associated risks and development plans of structural materials for in-vessel components (IVCs) in DEMO plants. Reduced Activation Ferritic Martensitic Steels form the primary structural materials for most DEMOs IVCs and will be the focus of the paper. An overview of EU- and J-DEMO programs RAFM steels developments is presented. We highlight the present status in their development for use in DEMO plants, with a focus on the structural materials' operational, and project-oriented, requirements. The development of materials property handbooks from high-quality data is illustrated. A process to validate these steels for operation in DEMO IVCs is summarised, revealing the pragmatic procedures ongoing, and limitations of this approach due to the synergistic operational effects on IVC materials; the use of in-situ or surveillance monitoring is outlined as a method to accommodate this limitation in synergistic effects and allow confidence in future DEMO operations. The development of advanced modifications to F82H and EUROFER are highlighted, with minor modifications leading to improved low and high-temperature operational design space being open for DEMO reactors. [ABSTRACT FROM AUTHOR]

DEMO will be a fusion power plant demonstrating the integration into the grid architecture of an electric utility grid. The design of the power conversion chain is of particular importance, as it must adequately account for the specifics of nuclear fusion on the generation side and ensure compatibility with the electric utility grid at all times. One of the special challenges the foreseen pulsed operation, which affects the operation of the entire heat transport chain. This requires a time-dependant analysis of different concept design approaches to ensure proof of reliable operation and efficiency to obtain nuclear licensing. Several architectures of Balance of Plant were conceived and developed during the DEMO Pre-Concept Design Phase in order to suit needs and constraints of the in-vessel systems, with particular regard to the different blanket concepts. At this early design stage, emphasis was given to the achievement of robust solutions for all essential Balance of Plant systems, which have chiefly to ensure feasible and flexible operation modes during the main DEMO operating phases – Pulse, Dwell and ramp-up/down – and to adsorb and compensate for potential fusion power fluctuations during plasma flat-top. Although some criticalities, requiring further design improvements were identified, these preliminary assessments showed that the investigated cooling system architectures have the capability to restore nominal conditions after any of the abovementioned cases and that the overall availability could meet the DEMO top-level requirements. This paper describes the results of the studies on the tokamak coolant and Power Conversion System (PCS) options and critically highlights the aspects that require further work. [ABSTRACT FROM AUTHOR]

• Application of Serpent 2 Monte Carlo code for nuclear fusion-relevant spectra and geometries. • Application of novel weight window methods in Serpent 2 to deep shielding problems • Benchmarking of Serpent 2 nuclear responses against SINBAD experimental data and MCNP. Released in 2009, the Serpent Monte Carlo code has established itself as a highly efficient and powerful simulation code for nuclear systems analysis. Originally developed for reactor physics applications, the scope of the code now extends to coupled multi-physics simulations and photon transport. The latter has allowed adoption of the code by the fusion neutronics community following developments of a coupled neutron-photon capability in 2014 and the ability to handle complex geometry types in 2016. The code is well validated for the energy regimes and geometry types one can expect in fission reactor analysis. Over the course of recent years a benchmarking effort has been undertaken for application of the code to nuclear fusion. Compared to nuclear fission, or accelerator based applications, the underlying particle interaction phenomena differ greatly at the energies expected in a fusion reactor as well as the specific responses that are of interest. In this paper, a novel weight window generation implementation in Serpent is investigated. The applicability of this method is demonstrated for the Frascati Neutron Generator (FNG) bulk blanket and shield experiment, part of the SINBAD database, and a DEMO helium cooled pebble bed (HCPB) computational model. A comparison is performed against MCNP using weight windows generated with ADVANTG. Excellent agreement is found for the specified tallies and the significant efficiency gain using weight windows generated using both methods is comparable. A robust variance reduction method implementation is fundamental to applications to fusion neutronics and as such, this work is an important step in deployment of Serpent for this type of analysis. [ABSTRACT FROM AUTHOR]

In a power plant scale fusion reactor, a huge amount of thermal power produced by the fusion reaction and external heating must be exhausted through the narrow area of the divertor targets. The targets must withstand the intense bombardment of the diverted particles where high heat fluxes are generated and erosion takes place on the surface. A considerable amount of volumetric nuclear heating power must also be exhausted. To cope with such an unprecedented power exhaust challenge, a highly efficient cooling capacity is required. Furthermore, the divertor must fulfill other critical functions such as nuclear shielding and channeling (and compression) of exhaust gas for pumping. Assuring the structural integrity of the neutron-irradiated (thus embrittled) components is a crucial prerequisite for a reliable operation over the lifetime. Safety, maintainability, availability, waste and costs are another points of consideration. In late 2020, the Pre-Conceptual Design activities to develop the divertor of the European demonstration fusion reactor were officially concluded. On this occasion, the baseline design and the key technology options were identified and verified by the project team (EUROfusion Work Package Divertor) based on seven years of R&D efforts and endorsed by Gate Review Panel. In this paper, an overview of the load specifications, brief descriptions of the design and the highlights of the technology R&D work are presented together with the further work still needed. [ABSTRACT FROM AUTHOR]

• An overview about the engineering aspects of the European DEMO wall protection strategy is given. • The background of the wall protection issue and design approaches are explained. • Recent outcomes from the interim design studies and technology R&D efforts are presented. • Focus is placed on the innovative technology achieved for a novel armor material and joining. • Results from the extreme and medium heat flux tests are also presented. Off-normal macroscopic plasma instabilities which possibly occur in a fusion power plant represent the most critical operational accidents and pose a serious risk to economically viable exploitation of fusion power. During such an instability, a huge amount of plasma energy is rapidly deposited onto the plasma-facing first wall in form of extreme heat flux. Such heavy thermal shocks can cause severe damages or a fatal failure of the first wall even by a single event. This issue is a serious design concern raising the requirement of a wall protection strategy. To cope with this issue, the European DEMO project adopted the limiter concept. The limiters equipped with tungsten sacrificial armor shall be installed at those wall areas where plasma contact is expected to be probable. To implement the limiter concept, comprehensive engineering studies have been conducted including conceptual design of the plasma facing component and technology R&D for armor material and joining. In this overview paper, the background of the wall protection issue and its implication on design and materials are explained and the outcomes from the recent project period (2019-20) are reported. The focus is placed on the technology achievement related to the development of novel lattice-type tungsten armor material fabricated by an innovative additive manufacturing process. Interim results from the extreme and medium heat flux tests and the joining trials for tungsten-to-heat sink material (copper or steel) are also addressed. The design rationale and the interpretation of the test data are supported by computational simulations. [ABSTRACT FROM AUTHOR]

This paper provides an overview of the methodology adopted within the EUROfusion framework covering the Pre-Concept Design Phase (2014–2020). The phase activities focused on advancing the technical basis of the design of the European DEMOnstration Fusion Power Plant (DEMO). The emphasis in the latter stages was placed upon an integrated design approach, based on systems engineering, providing a clear direction for urgent technology R&D and addressing the key design integration issues by assessing the critical systems interdependencies and uncertainties of certain design assumptions. The programme has continued to benefit from the ITER return of experience, in terms of design, licensing, and construction and DEMO will maintain a strong link with the ITER schedule in the coming design phases. The growing need for a structured and transparent review process gave rise to the introduction of a phase-gate review process, the implementation of which led to the first formal DEMO-wide gate review held at the end of the Pre-Concept Design Phase. The Gate Panel's recommendations provided clarity on the key priorities for the programme moving forward and serve to underpin the planning for the next phase. [ABSTRACT FROM AUTHOR]

An important issue for the Fusion Reactors is the hydrogen explosion hazard assessment. For this reason, in the Work Package of Safety Analyses and Environment (WPSAE) of the EUROFusion consortium, a task was established to identify the potential for hydrogen production in DEMO vessel in accident situations and to investigate the possible solutions which can prevent the risk of hydrogen and dust explosion. The hydrogen in the Vacuum Vessel (VV) can lead to combustion progressing in deflagration and detonation. Besides, the tungsten dust could enhance the effects of H 2 reaction as demonstrated by experiments, however, in this first set of accidents, only the H 2 generated by the chemical reaction is accounted for. For these reasons, the solutions to avoid or limit the H 2 production and/or to control safely the risk of H 2 accumulation and explosion need to be screened starting from the solutions adopted in the fission technology, and later on, adapted and assessed in the Pressure Suppression System (PSS) foreseen in DEMO. One promising idea seems to install passive catalytic recombiners in the PSS. The aim of this paper is firstly to set a range of operating parameters for the PAR functioning and intervention suitable in DEMO PSS configuration. Secondly to verify if the present criteria and parameters work efficiently in mitigating the H 2 accumulation a PSS in accidental conditions. To test the theoretical effectiveness of the PAR intervention, two reference accidents have been analyzed: an in-VV Loss Of Coolant Accident (LOCA) in the frame of the Design Basis Accident (DBA) and a Loss Of Flow Accident (LOFA) without active plasma shutdown in the context of the Beyond Design Basis Accidents (BDBA). The results seem promising if the performances of the PAR will be confirmed in sub atmospheric and saturated conditions. [ABSTRACT FROM AUTHOR]

• An overview of plant and tokamak building layout criteria of DEMO reactor are outlined. • A first preliminary layout of DEMO main systems is draft. • The main issues in arrangement systems in the tokamak building are pointed out. • The experience lesson learned from ITER are taken into account. • Possible solutions for each of those issues are briefly discussed. An early attention to the layout of both plant site and its buildings is essential in a complex plant under preliminary design as DEMO in order to meet the assigned targets, namely i) the licensing requirements ii) a good availability in delivery electricity to the grid. The layout definition has to follow several criteria that become more complex and stringent for nuclear buildings, e.g. functional, maintenance, fire protection, safety, human factors, shielding, and remote handling. The criterion As Low As Reasonable Achievable, with respect to the dose to the staff, has to be applied in design, operation, maintenance and decommissioning phases. The tokamak building, where several complex systems have to converge to the torus to create and control the plasma, to take out its energy and to produce and extract tritium, provides the second and ultimate confinement barrier between the environment and the hazardous and radioactive materials present inside that might be mobilised, in case of accident, by the high energetic fluids stored in DEMO systems. The layout criteria are focused on avoiding any challenge to the safety functions: e.g. no common mode failures of the safety classified systems for all reference design basis events. Furthermore the safety classified equipment have to maintain their safety function all over the plant life in such challenging environmental conditions; an accurate layout might allow the qualification possible, making milder the environment, e.g. defining adequate shielding and areas where radiation dose, magnetic field and accidental environmental conditions are reasonable for sensible equipment. The experience of NPPs and ITER is also recognized as the basis of such design criteria for DEMO. The paper will outline the main design basis events and the layout criteria presenting some applications for the tokamak building that reflect the recent progress of the DEMO design. [ABSTRACT FROM AUTHOR]

In the framework of the EU DEMO fusion reactor technology development, a new steam generator consisting in a helical tube bundle is currently under study for the lithium-lead loop of the Dual Coolant Lithium Lead (DCLL) and Water Cooled Lithium Lead (WCLL) breeding blankets. This solution turns out to be very interesting for high power plants, since the helical geometry is very compact and it assures a high thermal power exchanged, taking up the minimum amount of space. In this framework, the ENEA Brasimone Research Centre supports the development of this innovative component by means of experimental activities, involving CIRCE, a large scale pool-type facility using Lead-Bismuth Eutectic as primary fluid and pressurized water as secondary fluid. The main components of a new test section named THETIS are: a Fuel Pin Simulator, acting as primary heating source, a vertical mechanical pump for the primary fluid circulation inside the main pool and a new prototypical helical coil steam generator mock-up, designed to be relevant for the DCLL and WCLL lithium-lead loop, acting as primary heat sink. In such configuration, the facility will be involved in a set of tests aiming at demonstrating the technological feasibility and thermal-hydraulic performances of this prototypical steam generator, as well as the suitability of the component for the lithium-lead loop in DEMO. The experiments will also provide a database for system thermal-hydraulic codes validation. The aim of this paper is to present the main layout of the CIRCE facility, to describe the preliminary design of the test section and the main features of the helical coil steam generator mock-up. Furthermore, a preliminary test analysis carried out by the system thermal hydraulic code RELAP5/Mod3.3 is presented. A numerical 1-D model of the helical coil steam generator has been set-up in order to test the performance of the component from a thermal-hydraulic point of view. An additional 3D CFD analysis was performed for the flow field of the HCSG to assess the secondary flow behaviour in the component and the pressure losses. [ABSTRACT FROM AUTHOR]

In the frame of the activities promoted and encouraged by the EUROfusion Power Plant Physics and Technology (PPPT) department aimed at developing the EU-DEMO fusion reactor, strong emphasis has been recently addressed to the whole Balance of Plant (BoP) which represents the set of systems devoted to convert the plasma generated thermal power into electricity and to deliver it to the grid. Among these systems, the Divertor Primary Heat Transfer Systems (PHTSs) are intended to feed coolant to the two main components of the Divertor assembly, namely the Plasma Facing Components (PFCs) and the Cassette Body (CB). Since the DEMO Divertor must withstand high heat flux loads together with a considerable neutron deposited power, very tight tolerances may be allowed to the coolant inlet conditions. Therefore, the design of reliable PHTSs is of the utmost importance towards the development of an EU-DEMO fusion reactor. Within this framework, a study has been jointly carried out by University of Palermo, Ansaldo Nucleare and CREATE to design the Ex-Vessel PHTSs of both the PFCs and the CB for a DEMO reactor equipped with a Helium Cooled Pebble Bed Breeding Blanket. The paper describes criteria and rationale followed with the aim to achieve simple PHTS designs based on the adoption of easy-to-manufacture main components avoiding too much extrapolation from the state-of-art technology. Results of preliminary thermal-hydraulic calculations carried out to size heat exchangers, pressurizers, piping and pumps are presented and critically discussed, with particular attention to those integration, safety and feasibility constraints that may deeply affect the design of such components. Finally, the evaluation of PHTS key parameters as total pumping power and coolant inventory is reported. [ABSTRACT FROM AUTHOR]

In the framework of the DEMO Work Package Balance of Plant of EUROfusion consortium, ENEA has been involved in experimental and numerical activities related to the development of a prototypical heat exchanger, suitable as steam generator for the lithium-lead loop of the Dual Coolant Lithium Lead and Water Cooled Lithium Lead breeding blankets. For this purpose, an experimental campaign has been executed on the pool-type liquid metal-cooled facility CIRCE at ENEA Brasimone Research Centre. A dedicated test section named HERO has been designed and installed inside the main vessel of CIRCE. The innovative steam generator consists of a tube bundle with seven double walled bayonet tubes, fed with pressurized water. The selected configuration improves the plant safety, reducing the possibility of water-lead/lead-alloy interaction thanks to a double physical separation and allowing an easier control of eventual leakages from the coolant by pressurizing the separation region with inert gas. A set of tests has been defined to demonstrate technological feasibility and performances of this prototypical steam generator, as well as the suitability of the component for the lithium-lead loop in DEMO. In particular, one of the performed tests is presented and discussed in this paper. The experiment is characterized by a secondary loop feedwater working pressure of 10 MPa and a steam generator inlet temperature of 300 °C. On the primary side, the lead-bismuth eutectic has been used as working fluid with a steam generator inlet temperature of 480 °C. During the test, an experimental sensitivity analysis on the primary coolant mass flow rate has been performed. Furthermore, the results of a post-test analysis realized with two versions of the system thermal-hydraulic code RELAP5 are presented, in order to evaluate their capability in simulating the performances of the component and to support the validation process of the codes for heavy liquid metal applications. The work is concluded presenting a scaling analysis to find the equivalence between LBE and PbLi, recalculating the available experimental data with RELAP5 code using PbLi as working fluid. [ABSTRACT FROM AUTHOR]