ENEA Brasimone Research Centre Research Articles

Numerical study on benchmark analysis of NACIE-UP facility with uniform and non-uniform power distribution

The NACIE-UP (NAtural Circulation Experiment-Upgrade) facility is located at the ENEA Brasimone Research Centre (Italy) and is designed to investigate the thermal–hydraulic properties of Lead-Bismuth Eutectic (LBE) within a wire-spaced assembly under uniform and non-uniform power distribution conditions. This benchmark exercise proposed by ENEA is based on previous experiment results, intending for development and validation of CFD/TH system code, stand-alone CFD simulation, etc. In this study, the thermal–hydraulic characteristics of Fuel Pin Simulator (FPS) are analyzed, which contains inlet bend and outlet section. ANSYS Fluent solver is applied with the COUPLE algorithm of RANS SSTk-ω model. Cheng&Tak correlation and computational conditions are implanted into Fluent in the form of UDFs. Moreover, calculations were performed for ADP10, ADP06 and ADP07. The results show that the RMSEs of ADP10SS1 and ADP10SS2 are 2.52 K and 3.69 K, 7.05 K and 8.48 K for ADP06SS1 and ADP06SS2, respectively, which indicate that the adopted models have good prediction ability for the temperature of the experimental conditions. Meanwhile, the heat transfer characteristics of different sub-channels and sections are investigated and compared with the correlations of Ushakov, Mikityuk, Kazimi and Carelli. The experimental and simulation values of the uniformly heated ADP10 are in good agreement with the correlations, while there is a large deviation for the non-uniformly heated ADP06. Therefore, it is necessary to develop new correlations for the heat transfer characteristics of LBE in wire-wrapped bundle. Above all, this study provides quantitative and qualitative analysis for verification and validation of CFD/TH system code and stand-alone CFD simulation.

W-HYDRA: a new experimental platform for the Water-Cooled Lead Lithium Breeding Blanket

In the framework of the activities coordinated by the EUROfusion consortium, the Water thermal-HYDRAulic (W-HYDRA) experimental platform is being built at the ENEA Brasimone Research Centre in order to support the development of the Water-Cooled Lead Lithium (WCLL) Breeding Blanket (BB). In particular, this infrastructure will make possible the installation and testing of prototypical mock-ups under relevant working conditions, such as the First Wall (FW), the manifold and the Steam Generator (SG). Moreover, it will represent an integral test facility for the investigation of phenomena characteristic of WCLL BB concept, such as the PbLi/water interaction. Finally, the collection of data coming from the different planned experimental campaigns will allow to qualify and validate numerical models and codes currently adopted for the design of components, as well as for the modelling of complex phenomena typical of the WCLL BB. In order to come to a definitive design of the different facilities constituting the experimental platform, several design analyses assessing the thermal, hydraulic and structural performances of the different facilities and components are necessary. The paper reports a highlight of the W-HYDRA platform with a general description of the facilities. Some of the most relevant design studies carried out so far are reported as well, highlighting their impact on the evolution of the design.

Validation activities at ENEA Brasimone in support of the IFMIF-DONES design

IFMIF-DONES is a powerful neutron source which is being designed with the main purpose of qualifying structural materials (EUROFER being the first candidate) to be used in DEMO and fusion power plants envisaged after it. This source relies on one high current (125 mA) deuterons beam accelerated to 40 MeV which impacts on a liquid lithium target to produce an intense neutron flux through Li(d,n) stripping reactions able to simulate the nuclear responses expected on the first wall of the reactor. The engineering design of IFMIF-DONES is presently being carried out mainly in the framework of the EUROfusion Work Package Early Neutron Source (WPENS). Since 2021, a new phase has started with the launch of the FP9 WPENS workplan whose objective is to continue advancing the engineering design of the facility, putting special effort on the experimental validation of those aspects which still need to be qualified to demonstrate the fulfillment of functional and safety requirements. The ENEA Brasimone Research Centre has been and still is strongly committed in several validation tasks concerned in particular with the lithium systems design and the Remote Handling (RH) maintenance.In this paper, an overview of the most relevant validation activities carried out in recent years or still in progress or planned at the ENEA Brasimone Research Centre in both of the aforementioned areas is presented, including the erosion/corrosion tests in the Lifus 6 loop; the nitrogen-gettering materials qualification in the ANGEL facility; and the RH and prototypes testing in the DRP laboratory for the validation of the High Flux Test Module electric connectors and the pre-heating of the Target Assembly.

RELAP5/Mod3.3 thermal-hydraulics characterization of the steam generator mock-up during operational transients in STEAM facility in support of the design of the DEMO WCLL BoP

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.

Numerical analyses of the CIRCE-THETIS facility by mean of STH and CFD codes

During the last decade the European Union funded numerous projects with the aim to pave the way for the development of Gen IV nuclear reactor technologies. The recently started EU project PATRICIA, to which the University of Pisa participates, provides room for further investigations of the involved phenomena in LMFRs (e.g. thermal stratification inside the pool, LBE heat transfer capabilities, transition from forced to natural circulation) by means of both experimental and numerical analyses. In the frame of the PATRICIA project, the ENEA Brasimone Research Centre foresees to carry-out an experimental campaign involving the CIRCE facility. With respect to the previous configurations, the facility will be updated substituting the steam generator, which will exhibit an helicoidal steam generator (THETIS): the experimental campaign will involve the analysis of steady-state and postulated accidental scenarios. The University of Pisa provides numerical support both in the pre-test and post-test phases to assist the design of the experiment. This work reports about the preliminary results obtained by numerical simulations of the CIRCE-THETIS facility. The analyses were performed both using CFD codes (ANSYS Fluent) and STH codes (RELAP5-3D). The analyses mainly focused on temperature and velocity distributions inside the CIRCE pool during the postulated nominal steady-state conditions. In addition, a transient analysis was performed investigating the behaviour of the addressed facility in case of a PLOFA scenario. The limits and capabilities of both the approaches were observed and are discussed in the present work trying to provide guidelines for a correct application of the adopted codes. The obtained results provide interesting suggestions for the experimentalists and represent a valuable support in better setting the experimental conditions and measurements tools layout.

Experimental Results of PbLi-Water Reaction Performed in LIFUS5/Mod3 Separate Effect Test Facility

A major safety concern addressed during the design of the water-cooled lead-lithium (PbLi) breeding blanket (BB) is represented by an in-box loss-of-coolant accident, where high-pressure water is supposed to interact with PbLi inside the BB. Code development activities are being carried out to create the needed tools for the safety analysis of these systems in the case of incidental scenarios, which are supported by extensive experimental campaigns that aim at providing data for SIMMER code verification and validation (V&V). In this regard, the present work aims at presenting the dataset generated during the Series D experimental campaign performed at the LIFUS5/Mod3 facility operated at the ENEA Brasimone Research Centre. This is a separate effect facility able to simulate the mixing of water and PbLi alloy in conditions of temperature and pressure similar to the ones encountered by the system during nominal operation and to acquire significant data on all the relevant thermochemical parameters. Moreover, a preliminary analysis of the data has been performed to critically determine the quality of the data and to identify possible issues in the experimental process. In the end, the foreseen extension of the experimental work is described, as well as the foreseen application of the acquired data in the code V&V activities.

Liquid-Metal Thermal-Hydraulic Numerical Analyses in Support of the Upcoming CIRCE-THETIS Experimental Campaign

Liquid metal cooled reactors are among the design proposals accepted by the Generation IV International Forum for the fourth generation of nuclear power plants. During the last decade, many European Union (EU) projects started with the goal to pave the way for the development of this type of reactor. The present research work is performed in the framework of the EU Partitioning And Transmuter Research Initiative in a Collaborative Innovation Action (PATRICIA) project, supporting the development of the Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA). Among the cornerstones of the project are experimental and numerical analyses involving the CIRColazione Eutettico (CIRCE) facility set at the ENEA Brasimone Research Centre. The upcoming experimental campaign will address the new CIRCE-THETIS configuration, including a new kind of heat exchanger named THETIS: a helical coil steam generator (SG) whose heat transfer capabilities and impact on pool thermal hydraulics are to be investigated. The main goal of the present work is to provide numerical support to experimentalists to help them set up the CIRCE-THETIS experimental facility, the operating conditions test matrix, and postulated transients, providing information about potential needed updates or dangerous conditions. This paper reports on pretest computational fluid dynamics and system thermal-hydraulic analyses performed at the University of Pisa in the frame of the PATRICIA project. The new SG was first addressed assessing its heat transfer capabilities. Sensitivity analyses were performed; among the outcomes, the excellent heat transfer capabilities of the SG were highlighted. The pool and the reactor vessel auxiliary cooling system component were later investigated reporting valuable information concerning both postulated steady-state and transient conditions. The performed analyses had a relevant impact on the design of the facility, suggesting updates based on numerical calculations. After the experimental campaign, posttest analyses will be performed to draw lessons from the observed phenomena. This will also provide room for improvement of the adopted numerical tools.

Development of a coupling technique between RELAP5 and SIMMER-IV for fusion reactor applications

A critical problem in the Water-Cooled Lead-Lithium Breeding Blanket system (WCLL-BB) is the possible interaction between the water and the Lithium-Lead eutectic alloy - which act respectively as primary coolant and as breeder/neutron multiplier - due to a postulated rupture of the coolant circuit in the Breeding Unit of the BB. This scenario involves a complex multiphase interaction together with an exothermal chemical reaction between the two fluids with the production of hydrogen.The PbLi/water chemical reaction was implemented in SIMMER-IV code by the University of Pisa and, consequently, a coupling methodology was successfully developed between SIMMER-IV and RELAP5/Mod3.3 codes, in order to overcome SIMMER-IV unsuitability in the simulations of complex pipelines.This paper presents an application of the coupling methodology to the simulation of experimental tests, recently performed at ENEA inside the experimental campaign carried out with the LIFUS5/Mod3 facility at the ENEA Brasimone Research Centre. The injection line of the facility is simulated by RELAP5/Mod3.3, whilst the reaction vessel is simulated with SIMMER-IV.Results of different simulations are presented and compared against experimental data, providing both qualitative and quantitative evaluations of the performance of the coupling methodology in the prediction of the chemical and thermal-hydraulic phenomena involved in the experiments, such as the fast pressurisation of the injection line and the pressurisation of the reaction vessel, the energy release due to the chemical reaction and the propagation of pressure waves inside the reaction vessel.

Lithium-Lead/water interaction: LIFUS5/Mod3 series E tests analysed by SIMMER-III coupled with RELAP5

The Breeding Blanket is a necessary component to close the nuclear fusion reactor fuel cycle. amongst the most promising conceptual design, there is the Water Cooled Lithium Lead Breeding Blanket, with water as coolant and eutectic Lithium-Lead as neutron multiplier and breeder. The possible interaction between water and Lithium-Lead poses a main safety concern and prompted the scientific community to develop a numerical analysis tool able to simulate such a complex interaction. The SIMMER-III code was modified by UNIPI to simulate the chemical interaction between water and Lithium-Lead, furthermore also a coupling methodology between SIMMER-III and RELAP5/Mod3.3 was developed. The coupling tool employs SIMMER-III code to simulate the zone of Lithium-Lead interacting with water, whilst the RELAP5 code is used to simulate the water pipelines. The LIFUS5/Mod3 facility at the ENEA Brasimone Research Centre was designed to perform reliable experimental activities on the interaction between water and Lithium-Lead. In this facility water at high pressure is injected inside a reaction vessel, where the thermodynamic and chemical interaction between water and Lithium-Lead occurs. The experimental activities are divided in different tests matrix, the Series D and Series E tests. The two series differ in the amount of water injected during the transient. In series D the mass of water is predetermined whilst in series E water was injected continuously for a pre-fixed interval of time and the total injected mass was estimated a posteriori. This work shows the results of the coupling tool applied to Series E. The comparison between the experimental and numerical results is performed by identifying and characterizing the phenomena involved in the interaction. Furthermore, the overall performance of the coupled codes in the simulation of the phenomena is presented here.

SORGENTINA-RF FUSION NEUTRON PLANT: PRELIMINARY DESIGN OF THE BIOSHIELDING IN COMPLIANCE WITH DOSE CONSTRAINTS FOR WORKERS EXPOSURE.

An accelerator-driven 14MeV neutron source of new concept, denominated SORGENTINA-RF, will be installed in ENEA Brasimone Research Centre, to test the feasibility of producing radionuclides of medical relevance using fusion neutrons. The main goal of the facility is generating 99Mo as precursor of 99mTc, a radionuclide widely used in nuclear medicine diagnostic procedures, using the 14MeV fusion neutrons produced by the plant. This work describes the study performed for the design of a proper shielding structure that aims at fulfilling the requirement of 0.01mSv/h dose rate limit on the external surface of the shielding during beam-on operations. The proposed shielding consists of a layered structure composed of 2m standard concrete and 1m baritic concrete. The design is still in the preliminary phase to assess the feasibility and the economic issues as well as structural impact of the shielding structure.

Compatibility of niobium, vanadium and P22 steel in high temperature flowing LiPb

The LiPb loops of the WCLL BB (Water Cooled Lithium-Lead Breeding Blanket) have the primary aim of routing tritium-rich LiPb from the Breeding Blanket to the Tritium Extraction System of DEMO Reactor. Besides, the loops maintain LiPb in the desired conditions and purify it from corrosion and activated products and from helium generated by the neutron transmutation of lithium. For some of these tasks the loops need special functional materials. Among them, niobium and vanadium are widely considered for the tritium extraction technologies for their high permeability to tritium. In addition, the loops will be made of structural materials capable of withstanding the operative conditions and the aggressive environment. The current reference material for the piping is the ASTM A335 Grade P22. This paper investigates the compatibility at 500 °C of niobium, vanadium and P22 with LiPb flowing at 0.5 m/s. This velocity is the maximum foreseen in the LiPb loops, while the temperature was chosen following a conservative approach, as the maximum expected temperature is 450 °C in the BB and 330 °C in the LiPb loops. The specimens of the three materials were exposed for 1,000, 2,000 and 4,000 h in IELLLO LiPb facility at ENEA Brasimone Research centre. After the exposure, the specimens were examined by SEM-EDX, both in cross-section and surface view, XRD analyses were performed in order to investigate potential corrosion compounds formed. Weight loss analyses were also performed on niobium specimens, after chemical cleaning from adherent LiPb. The analyses highlighted the impact of corrosion products and their interaction with refractory metals (Ni and Fe) and P22 steel (Cr, after redeposition).

TRIEX-II: an experimental facility for the characterization of the tritium extraction unit of the WCLL blanket of ITER and DEMO fusion reactors

The experimental qualification of the tritium extraction unit (TEU) from the LiPb eutectic alloy (15.7 at. % Li), the breeder material of the water-cooled lithium–lead breeding blanket concept, is one of the fundamental items for the demonstration of tritium balance sustainability for ITER and DEMO fusion reactors. Several technologies have been proposed as TEU, but the selection of the reference technology can be carried out only after the experimental measurement of the tritium extraction efficiency. For this purpose, a dedicated facility, called TRIEX-II, was designed and installed at ENEA Brasimone Research Centre, Italy. The facility is able to qualify gas–liquid contactor (GLC), permeator against vacuum and liquid-vacuum contactor technologies at different temperatures, lithium–lead mass flow rates and hydrogen isotopes concentrations. In TRIEX-II, the hydrogen or deuterium, used to simulate tritium, are solubilised inside the LiPb with a dedicated saturator and are then extracted from the liquid metal in the GLC mock-up, which uses pure helium or a mixture constituted by helium and hydrogen as stripping gas and works in the temperature range between 300 and 500 °C.

HyPer-QuarCh II: A laboratory-scale device for hydrogen isotopes permeation experiments

In a D-T fusion reactor, the correct estimation of the tritium inventory and permeation fluxes towards the coolant and to the external environment is a crucial issue for the reactor licensing. Within this frame, a fast and reliable sensor for the online measurement of hydrogen isotopes concentration in the breeder is therefore necessary. At ENEA Brasimone research centre, Italy, the development, qualification and characterization of hydrogen isotopes permeation sensors (HPS) were carried out since the early 2000s. A new experimental laboratory-scale device, named Hyper-Quarch II (Hydrogen Permeation Quartz Chamber), was developed on the basis of the experience gained in the past experimental campaigns. This device is characterised by an upgraded test section in quartz and new instrumentation equipment, and will be used to test advanced hydrogen permeation sensor in both gas phase and in stagnant LiPb eutectic alloy (15.7 at.% Li). Hydrogen or deuterium will be adopted instead of tritium to simulate the operative conditions of the Water-Cooled Lithium-Lead (WCLL) Test Blanket Module of ITER and the WCLL Breeding Blanket of the European DEMO reactor. Moreover, the upgrade was performed to also allow the measurement of the Sieverts’ constant of hydrogen or deuterium solubilised in the LiPb with absorption or desorption techniques in a temperature range from 300 to 550∘C and pressure range 0.1 to 100 hPa.

Preliminary design of a helical coil steam generator mock-up for the CIRCE facility for the development of DEMO LiPb heat exchanger

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.

Experimental Investigation on CIRCE-HERO for the EU DEMO PbLi/Water Heat Exchanger Development

The present paper describes the experimental campaign executed at the ENEA Brasimone Research Centre aiming at supporting the development of a PbLi/water heat exchanger suitable for the lithium–lead loops of the dual coolant lithium lead and the water cooled lithium lead breeding blankets of the EU DEMO fusion reactor. The experiments were performed in a test section named HERO, installed inside the main vessel of the lead–bismuth eutectic-cooled pool-type facility CIRCE. The test section hosts a steam generator bayonet tube mock-up in relevant scale, which was selected as a promising configuration for DEMO purposes. For the thermal-hydraulic characterization of the component, five tests were executed at different water pressures (6, 8, 12 MPa, two tests at 10 MPa), and liquid metal flow rates (40, 33, 27, 20, 10 kg/s). The experimental outcomes proved the technological feasibility of this novel steam generator and its suitability for the DEMO PbLi loops. The activity was completed with a post-test analysis using two versions of the system code RELAP5. Because the experiments were executed with lead–bismuth eutectic, a scaling analysis is proposed to find the equivalence with PbLi. RELAP5 code was applied to recalculate the experimental data using PbLi as working fluid.

LIFUS II corrosion loop final design and screening of an Al based diffusion coating in stagnant LLE environment

Corrosion phenomena of structural material in Lead-Lithium Eutectic (LLE) environment is influenced by the coolant chemistry, temperature, velocity profile and impurities concentration dissolved in it. In the framework of the EUROfusion consortium, the Experimental Engineering Division (FSN-ING) by the ENEA Brasimone Research Centre (RC) is developing scientific activities to investigate the corrosion rate of materials and to test Al based diffusion coating at 550 °C, different velocities (0.01, 0.1 and 1 m/s) and different exposure times. To this end, a new experimental facility, named LIFUS II (Lithium for Fusion II) is under construction at the Brasimone RC laboratories. Piping and components installed in the hot leg of the facility are internally coated by an Al based diffusion coating developed in collaboration with RINA Consulting-CSM to reduce concentration of corrosion product in the loop. A preliminary screening of the coating was performed in LLE stagnant conditions at 550 °C for 1000 h showing a good resistance and no signs of dissolution were detected. The only modification on the surface was found to be related to the formation of dark islands made of Al2O3.

Fusion technologies development at ENEA Brasimone Research Centre: Status and perspectives

In the European and international framework, ENEA coordinates the Italian fusion program, supported by linked third parties as universities, research institutes and industries. In this context, the Experimental Engineering Division (FSN-ING) is involved in experimental and numerical activities related to Breeding Blanket (BB) and Divertor technologies development.This paper describes the scientific works presently ongoing at Brasimone R.C. enveloped in LLE, lithium, helium, tritium and pressurized water technologies, characterization of structural materials, analysis of materials corrosion rate and development and qualification of anti-permeation/corrosion barrier. The experimental activities conducted for the investigation of safety-relevant scenarios as In-Box LOCA (LLE-water interaction in the WCLL or shock waves propagation generated by helium injection in LLE in the HCLL/DCLL-BBs) are here reported.Finally, new activities have been planned to support the Divertor Tokamak Test divertor characterization, the large-scale LLE-water interaction and a LLE components validation in relevant scale for WCLL-BB.

An integrated hydrogen isotopes transport model for the TRIEX-II facility

Abstract At ENEA Brasimone Research Centre, Italy, a new experimental facility named TRIEX-II ( Tri tium Ex traction) was designed and installed. Its aim is to characterize, in the range of operating conditions foreseen for the European Test Blanket System WCLL-TBS, several extraction technologies for hydrogen isotopes (Q2) solubilized in the flowing metallic LiPb alloy (15.7 at. % Li). One of these technologies is the packed column, an example of Gas/Liquid Contactors (GLCs). This paper proposes a multiscale modelling tool, combining different scales through two computational tools. The extraction column mock-up is described by a component-detail level model, developed in COMSOL Multiphysics, and integrated into a system level code of the whole TRIEX-II circuit, developed using MATLAB/Simulink. The integration is carried out by implementing the COMSOL component into an S-function of MATLAB/Simulink, preserving the process flow diagram of the loop. In this way, it was possible to quantify the Q2 concentrations and the permeation fluxes, and to evaluate the theoretical extraction efficiency. The LiPb flow field inside the extractor was also derived. Finally, a comparison with the experimental results was performed. The results suggest that this tool could be adapted for analyses of complex systems, at a multiscale level, in view of design improvements and safety studies for the tritium cycle of ITER.

Simulation of operational conditions of HX-HERO in the CIRCE facility with CFD/STH coupled codes

The paper describes the application of a coupled methodology between Fluent CFD code and RELAP5 System Thermal-Hydraulic code developed at the DICI (Dipartimento di Ingegneria Civile e Industriale) of the University of Pisa. The methodology was applied specifically to the LBE-water heat exchanger HERO located inside the S100 vessel of the CIRCE facility, built at ENEA Brasimone Research Centre, to investigate the capabilities of this component. In the proposed methodology, the primary side of the HX-HERO, containing LBE, is simulated by the CFD code, while the secondary side, containing a two phase mixture of water and vapour, is reproduced by the System Thermal-Hydraulic code. During the calculation the two codes exchange, at the coupled boundaries: the bulk temperature and heat transfer coefficient of the ascending water (RELAP5 to Fluent) and the wall temperature at the water side surface of the pipes (Fluent to RELAP5). The coupling technique was tested by comparing the numerical results with the experimental data recently obtained by ENEA; the numerical results predicted well the qualitative trend of the temperature and provided an overall good prediction of the temperature also from a quantitative point of view. It is worth noticing that this good performance remained reliable for all the cases simulated, proving the general applicability of the methodology.

Post-test simulations for the NACIE-UP benchmark by STH codes

This paper illustrates the results obtained in the last phase of the NACIE-UP benchmark activity foreseen inside the EU SESAME Project. The purpose of this research activity, performed by system thermal–hydraulic (STH) codes, is finalized to the improvement, development and validation of existing STH codes for Heavy Liquid Metal (HLM) systems. All the participants improved their modelling of the NACIE-UP facility, respect to the initial blind simulation phase, adopting the actual experimental boundary conditions and reducing as much as possible sources of uncertainty in their numerical model. Four different STH codes were employed by the participants to the benchmark to model the NACIE-UP facility, namely: CATHARE for ENEA, ATHLET for GRS, RELAP5-3D© for the “Sapienza” University of Rome and RELAP5/Mod3.3(modified) for the University of Pisa. Three reference tests foreseen in the NACIE-UP benchmark and carried out at ENEA Brasimone Research Centre were analysed from four participants.The data from the post-test analyses, performed independently by the participant using different STH codes, were compared together and with the available experimental results and critically discussed.