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The water-cooled lithium–lead breeding blanket is a candidate option for the European Demonstration Power Plant(DEMO) nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium– lead as breeder–multiplier,pressurized water as coolant, and EUROFER as structural material. The current design is based on DEMO 2015specifications and represents the follow-up of the design developed in 2015. The single-module-segment approach isemployed. This is constituted by a basic geometry repeated along the poloidal direction. The power is removed by meansof radial–toroidal (i.e., horizontal) water cooling tubes in the breeding zone. The lithium–lead flows in a radial–poloidaldirection. On the back of the segment, a 100-mm-thick plate is in charge of withstanding the loads due to normal operationand selected postulated initiating events. Water and lithium–lead manifolds are designed and integrated with a consistentprimary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium–leadsystem. Rationale and features of the single-module-segment water-cooled lithium–lead breeding blanket design arediscussed and supported by thermo-mechanic, thermo-hydraulic, and neutronic analyses. Preliminary integration withthe primary heat transfer system, the energy storage system, and the balance of plant is briefly discussed. Open issues, areasof research, and development needs are finally pointed out.

The water-cooled lithium-lead breeding blanket is a candidate option for the European Demonstration Power Plant (DEMO) nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium-lead as breeder-multiplier, pressurized water as coolant, and EUROFER as structural material. The current design is based on DEMO 2015 specifications and represents the follow-up of the design developed in 2015. The single-module-segment approach is employed. This is constituted by a basic geometry repeated along the poloidal direction. The power is removed by means of radial-toroidal (i.e., horizontal) water cooling tubes in the breeding zone. The lithium-lead flows in a radial-poloidal direction. On the back of the segment, a 100-mm-thick plate is in charge of withstanding the loads due to normal operation and selected postulated initiating events. Water and lithium-lead manifolds are designed and integrated with a consistent primary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium-lead system. Rationale and features of the single-module-segment water-cooled lithium-lead breeding blanket design are discussed and supported by thermo-mechanic, thermo-hydraulic, and neutronic analyses. Preliminary integration with the primary heat transfer system, the energy storage system, and the balance of plant is briefly discussed. Open issues, areas of research, and development needs are finally pointed out. @EUROfusion Consortium

As part of a Department of Energy International Nuclear Energy Research Initiative (1-NERI), the Dutch Nuclear Research and consultancy Group (NRG), the Belgian Nuclear Research Centre (SCKCEN), Ghent University (UGent) in Belgium and the Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) are collaborating with Argonne National Laboratory (ANL) to perform and compare a series of thermal hydraulic simulations representative of a heavy liquid metal fast reactor fuel assembly. Such a widely spaced wire-wrapped fuel assembly is a complex configuration for which few flow data are available for verification and validation of computational fluid dynamics (CFD) simulations. For this benchmark a 19-pin wire-wrapped rod bundle with characteristics representative of the MYRRHA flexible fast research reactor, under design at SCK'CEN in Belgium, is modeled. The heat conduction in the cladding of the fuel rods and the spacer wires is taken into account by conjugate heat transfer. UGent, ENEA and NRG performed their Reynolds Averaged Navier-Stokes (RANS) simulations with commercially available CFD codes. The high-fidelity ANL Large-Eddy Simulation (LES) was performed with Nek5000, used for CFD in the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite. The paper will show and discuss the comparison of the thermal and hydraulic RANS results and the reference Nek5000 LES results. The comparison with the LES results will indicate to which extent the current liquid metal modeling methods are sufficient and help to highlight remaining issues. The results of the study are very valuable in the design and licensing process for MYRRHA. © Copyright (2015) by American Nuclear Society All rights reserved.

A detailed study of lithium-related topics in the IFMIF-DONES facility is currently being promoted and supported within the EUROfusion action, paying attention to different pivotal aspects including lithium flow stability and the monitoring and extraction of impurities. The resistivity meter is a device able to monitor online non-metallic impurities (mainly nitrogen) in flowing lithium. It relies on the variation of the electric resistivity produced by dissolved anions: the higher the concentration of impurities in lithium, the higher the resistivity measured. The current configuration of the resistivity meter has shown different measuring issues during its operation. All these issues reduce the accuracy of the measurements performed with this instrument and introduce relevant noise affecting the resistance value. This paper proposes different upgrades, supported by CFD simulations, to optimize lithium flow conditions and to reduce measurement problems. Owing to these upgrades, a new design of the resistivity meter has been achieved, which is simpler and easier to manufacture.

The SORGENTINA-RF project is presented in terms of general structure and description of the main tasks and activities to be carried out. It is devoted to the design and development of a medium power 14 MeV fusion neutron source relying on a rotating target and a deuterium/tritium ion accelerator. The main focus of the neutron facility is the production of radiopharmaceutical precursors, in particular 99Mo as precursor of 99mTc, a radio-tracer used in single photon emission computed tomography. The nuclear reaction involved in the production of 99Mo is the inelastic reaction 100Mo(n,2n)99Mo. The facility will assess the chain that starts with the irradiation of the natural molybdenum (where 100Mo has an isotopic abundance of about 10%) up to the production of the so-called mother solution, a liquid solution named sodium molybdate. The facility will also make available fast and thermal neutrons beams for studies on innovative medical radioisotopes as well as materials.

In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the coolability of the Fuel Assembly in nominal condition is of central interest. The Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED) is a 300 MWth pool-type reactor aimed at demonstrating the safe deployment of the Generation IV LFR technology. The ALFRED design, currently being developed by the Fostering ALFRED Construction international consortium, is based on prototypical solutions intended to be used in the next generation of lead-cooled Small Modular Reactors. Within the scope of FALCON and in the frame of investigating the thermal-hydraulics of the ALFRED core, a CFD computational model of the general Fuel Assembly (FA) is built looking for the assessment of its thermal field in nominal flow conditions both for the average FA and the hottest one. Starting from the experience in this kind of simulations and in experimental work, the whole model of the ALFRED Fuel Assembly is first presented and calculation of flow and temperature field in nominal conditions is carried out. Results showed that the thermal hydraulic field predicted in the average FA by the code is in good agreement with analytical correlations and the temperature field on the pin clad is acceptable for clad material temperature constraint. About the results on the hot FA test case, the CFD results highlighted a peak temperature on the clad close to the clad temperature constraint. This result led to an upgrade of the mass flow distribution among the FA for achieving a 20% mass flow increase in the hottest one that guarantees higher temperature margin on the clad.

The NEXTOWER H2020 EU project investigates the possibility of using liquid lead as heat storage medium for high-temperature Thermal Energy Storage (TES) in concentrated solar power plants. To that end, within such project, a demonstration TES unit named SOLEAD (SOlar LEAd Demonstrator) is being developed to be coupled with an open volumetric air receiver in a solar tower CSP system. The SOLEAD demonstrator will use pure lead as working fluid and will be tested stand-alone, to address structural material behaviour at very high temperature, without coupling with air receiver. The tests are planned by the ENEA Brasimone R.C. in Italy. The introduction of the paper provides a general summary of the TES and CSP technology. Then a section is devoted to the conceptual design of SOLEAD. In the lead stand-alone experiment, the focus is on the materials corrosion in lead environment. Although the thermal stratification in the main vessel cannot be reproduced in a stand-alone experiment, the thermal cycle of the facility in the range 600-750°C will be properly reproduced in the experimental test. To this aim, external heating cables will heat up the system from 600°C to 750°C, while a proper Air Cooling System (ACS) will cool down the lead pool. A section of the paper contains the design criteria and calculation of the ACS with the air flowing on an annular gap between the vessel and the insulation. The power provided and extracted by the two systems (heating cables and ACS) is around 30 kW, so that the temperature range 600-750°C can be covered in about 8 hours and a complete cycle can be carried out in one day. Finally, a brief summary of the operational procedures needed from lead melting to the materials inspection is provided. The plan is to test the vessel for 4 months with daily thermal cycles to assess the resistance of FeCrAl materials exposed to very high temperature lead.

One of the requirements for achieving high levels of safety in fourth generation nuclear power plants, is that core thermal-hydraulics can be simulated numerically at a good level of accuracy. To this aim, detailed validation of turbulent heat transfer models needs to be carried out and discussed, in relevant flow and heat transfer configurations, and for low Prandtl-number fluids. The focus of this research is the turbulent, fully developed convection in a heated rod bundle with a triangular arrangement, and a pitch-to-diameter ratio P / D = 1.4 . Statistics derived from a set of Direct Numerical Simulations (DNS) at the moderate Reynolds number of Re = 8290 are compared to Reynolds Averaged Navier-Stokes (RANS) solutions where the closure is provided by the two-equation model SST k- ω and by the Reynolds-Stress Model. Comparisons are drawn for forced flow and in mixed convection conditions ( Ri ≈ 0.25 ). Global quantities extracted from experiments performed in very similar conditions are also compared against the numerical results. Profiles of the turbulent Prandtl number about the unit flow cell are also displayed and discussed. This work clarifies through detailed comparison that if, on the one hand, only the Reynolds-Stress model is able to reasonably capture important flow features like secondary flow components, on the other hand also the SST k- ω two-equation model considered is acceptably accurate in predicting the integral quantities of interest.

An experimental campaign was performed on a non-uniformly heated 19-pins wire-spaced fuel pin bundle simulator, cooled by Heavy Liquid Metal and installed in the NACIE-UP (NAtural CIrculation Experiment-UPgrade) facility located at the ENEA Brasimone Research Center (Italy). The experimental tests concerned mass flow rate transition of the primary coolant from forced to natural circulation, with fuel pin bundle simulator characterized by non-uniform power distribution. The main objective of the experimental campaign was to perform integral system and local thermal-hydraulic analysis, investigate different flow regimes and the transition from forced to natural circulation flow and, more specifically, analyze the behavior of the 19-pins wire-spaced Fuel Pin Simulator (FPS) during such transient. Indeed, the performed test were characterized by non-uniform heating of the bundle (i.e. just some pins switched on), so the effects of this non-uniformity on the local temperatures and on the overall system behavior was evaluated. A deep investigation on the local temperature distribution was performed thanks to the accurate instrumentation provided in the bundle (67 thermocouples). The obtained experimental data provided useful information for the characterization of the bundle and the computation of the heat transfer coefficient. A post-test validation is made by a detailed CFD model of the test section. Three monitored section at different levels are compared both for wall and bulk temperatures. This post-test comparison with this experimental configuration is unique and represents a further step towards the validation of the CFD models and methods in fuel bundle geometries cooled by HLM.

In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the flow blockage in a fuel sub-assembly is considered one of the main issues to be addressed and one of the most important and realistic accident for Lead Fast Reactors (LFR) fuel assembly. The blockage in a fast reactor Fuel Assembly (FA) may have serious effects on the safety of the plant leading to the FA damaging or melting. The external or internal blockage of the FA may impair the correct cooling of the fuel pins, be the root cause of anomalous heating of the cladding and of the wrapper and potentially impact also fuel pins not directly located above or around the blocked area. In order to model the temperature and velocity field inside a wrapped FA under unblocked and blocked conditions, detailed CFD thermal hydraulic analyses of the FA are required. ALFRED is a 300 MWth Lead-cooled fast reactor GEN.IV concept. The advanced conceptual design of the reactor was carried out within the LEADER EU project in the last years. In the SESAME EU project specific thermal-hydraulic experiments and simulations were addressed to explore the basic phenomenology of some accidental events like internal flow blockage. A CFD computational model of the ALFRED FA is built for thermal hydraulic analysis of relevant internal blockage scenarios. The whole model of the ALFRED Fuel Assembly is first presented and calculation of flow and temperature field in nominal condition is carried out. RANS simulations of idealized blockage scenarios like one sector and two-sectors blockage are performed adopting three different spacer grid location (under the active length, at half active length, above the active length). Results showed that the most likely blockage in the lower grid positioned before the active region do not perturb the temperature distribution in the fuel assembly, while the central grid ones have strong consequences and leads to a clad temperature peak behind the blockage with possible clad failure. On the other hand, the most severe case on the upper grid (2 sectors blockage) showed a manageable temperature maximum (700°C) at the end of the active region due to the lower mass flow rate in the FA. As a conclusion lower and upper grid blockages perturb only marginally the temperature field and do not lead to serious clad failure involving only creep long-term effects, while a blockage in the active region leads to serious clad damage and possible clad deformation and fusion

The next generation of nuclear energy systems, also known as Generation IV reactors, are being developed to meet the highest targets of safety and reliability, sustainability, economics, proliferation resistance, and physical protection, with improved performances compared with the currently licensed plants or those presently being built. Among the proposed technologies, lead-cooled fast reactors (LFRs) have been identified by nuclear industries in both Western and developing countries as being among the optimal Generation IV candidates. Since 2000, ENEA, the Italian National Agency for New Technologies, Energy, and Sustainable Economic Development is supporting the core design, safety assessment, and technological development of innovative nuclear systems cooled by heavy liquid metals (HLM) and, most recently, fully oriented on LFRs. ENEA is developing world-recognized skills in fast spectrum core design and is one of the largest European fleets of experimental facilities aiming at investigating HLM thermal-hydraulics, coolant chemistry control, corrosion behavior for structural materials, and material properties in the HLM environment, as well as at developing corrosion-protective coatings, components, instrumentation, and innovative systems, supported by experiments and numerical tools. Efforts are also dedicated to develop and validate numerical tools for specific application to HLM systems, ranging from neutronics codes, system and core thermal-hydraulic codes, computational fluid dynamics (CFD) and fuel pin performance codes, including their coupling. The present work aims at highlighting the capabilities and competencies developed by ENEA so far in the framework of liquid metal technologies for Generation IV LFRs. In particular, an overview on the ongoing R&D experimental program will be depicted considering the current fleet of facilities, namely: CIRCE, NACIE-UP, LIFUS5, LECOR, BID-1, HELENA, RACHEL, and Mechanical Labs. An overview on the numerical activities performed so far and those presently ongoing is also reported. Finally, an overview of the ENEA contribution to the ALFRED Project in the frame of the FALCON international consortium is reported, mainly addressing the ongoing activity in terms of core design, technology development, and auxiliary systems design.