GEN-IV LFR development: Status & perspectives

Since Lead-cooled Fast Reactors (LFR) have been conceptualized in the frame of Generation IV International Forum (GIF), great interest has focused on the development and testing of new technologies related to Heavy Liquid Metal (HLM) nuclear reactors. In this frame, ENEA developed one of the larger European experimental fleet of experimental facilities aiming at investigating HLM thermal-hydraulics, coolant chemistry control, corrosion behavior for structural materials, and at developing components, instrumentations and innovative systems, supported by experiments and numerical tools. The present work aims at highlighting the capabilities and competencies developed by ENEA so far in the frame of the liquid metal technologies for GEN-IV LFR. In particular, an overview on the ongoing R&D experimental program will be depicted considering the actual fleet of facilities: CIRCE, NACIE-UP, LIFUS5, LECOR and HELENA. CIRCE (CIRColazione Eutettico) is the largest HLM pool facility presently in operation worldwide. Full scale component tests, thermal stratification studies, operational and accidental transients and integral tests for the nuclear safety and SGTR (Steam Generator Tube Rupture) events in a large pool system can be studied. NACIE-UP (NAtural CIrculation Experiment-UPgraded) is a loop with a HLM primary and pressurized water secondary side and a 250 kW power Fuel Pin Simulator working in natural and mixed convection. LIFUS5 (lithium for fusion) is a separated effect facility devoted to the HLM/Water interaction. HELENA (HEavy Liquid metal Experimental loop for advanced Nuclear applications) is a pure lead loop with a mechanical pump for high flow rates experiments. LECOR (LEad CORrosion) is a corrosion loop facility with oxygen control system installed. All the experiment actually ongoing on these facilities are described in the paper, depicting their role in the context of GEN-IV LFR development

Conceptual Design of the Steam Generators for the EU DEMO WCLL Reactor

In the framework of the EUROfusion Horizon Europe Programme, ENEA and its linked third parties are in charge of the conceptual design of the steam generators belonging to EU DEMO WCLL Breeding Blanket Primary Heat Transfer Systems (BB PHTSs). In particular, in 2021, design activities and supporting numerical simulations were carried out in order to achieve a feasible and robust preliminary concept design of the Once Through Steam Generators (OTSGs), selected as reference technology for the DEMO Balance of Plant at the end of the Horizon 2020 Programme. The design of these components is very challenging. In fact, the steam generators have to deliver the thermal power removed from the two principal blanket subsystems, i.e., the First Wall (FW) and the Breeding Zone (BZ), to the Power Conversion System (PCS) for its conversion into electricity, operating under plasma pulsed regime and staying in dwell period at a very low power level (decay power). Consequently, the OTSG stability and control represent a key point for these systems' operability and the success of a DEMO BoP configuration with direct coupling between the BB PHTS and the PCS. In this paper, the authors reported and critically discussed the FW and BZ steam generators' thermal-hydraulic and mechanical design, the developed 3D CAD models, as well as the main results of the stability analyses and the control strategy to be adopted

Experimental and numerical analysis of heavy liquid metal systems for Generation IV fast reactors

The present work reports the results achieved during the doctoral research activity realized in partnership between DIAEE (Dipartimento di Ingegneria Astronautica, Elettrica ed Energetica) of Sapienza University of Rome and ENEA (Agenzia nazionale per le nuove tecnologie, l’energia e lo sviluppo economico sostenibile). The activities have been carried out within the EU scientific community, since they are part of the R&D activities foreseen in the two HORIZON2020 European Projects SESAME and MYRTE. After a brief description of the Lead Fast Reactors (LFRs) technologies and the actual status of the related R&D programs worldwide, the description of the Lead Bismuth Eutectic (LBE)-cooled pool-type facility CIRCE is presented. In particular, the work is focalized to the newest test section presently installed on CIRCE and named HERO (Heavy liquid mEtal pRessurized water cOoled tubes). The performed experimental campaigns aimed at characterizing a prototypical steam generator with double-wall bayonet tubes, evaluating its thermal-hydraulic performances in normal operational and transient scenarios. The experimental activity on CIRCE-HERO has been supported by a numerical pre-test analysis described in the third section of this document. In particular, the RELAP5-3D© model of the HERO secondary loop has been set-up and it has been used to define the start-up procedure of the facility and to achieve feedbacks on the performances of the steam generator. The core of this document is dedicated to the description and post-test analysis of the two experimental campaigns executed on CIRCE-HERO. The first experimental campaign, consisting of three tests, has been performed in the framework of the HORIZON2020 SESAME EU project, with the objective to support the development of the ALFRED design. The second one, consisting of nine tests, has been executed in the framework of the HORIZON2020 MYRTE EU project, with the purpose to support the development of MYRRHA and acquiring experimental data relevant for MYRRHA primary heat exchanger. To extend the knowledge and validation of SYS-TH codes when applied for LFRs, a simulation activity has been performed in the Benchmark exercise for SYStem Thermal-Hydraulic (SYS-TH) codes and CFD/SYS-TH codes validation, in the framework of the H2020 SESAME project. A RELAP5-3D© model of the NACIE-UP facility has been set up and it has been involved to perform a preliminary blind simulation activity and a subsequent post-test analysis on the basis of the experimental results available from the test performed on NACIE-UP. A final summary, conclusions and future perspectives are given in the final section of the document

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

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

Development of a Steam Generator Mock-Up for EU DEMO Fusion Reactor: Conceptual Design and Code Assessment

Recent R&D activities in nuclear fusion have identified the DEMO reactor as the ITER successor, aiming at demonstrating the technical feasibility of fusion plants, along with their commercial exploitation. However, the pulsed operation of the machine causes an “unconventional” operation of the system, posing unique challenges to the functional feasibility of the steam generator, for which it is necessary to define and qualify a reference configuration for DEMO. In order to facilitate the transitions between different operational regimes, the Once Through Steam Generator (OTSG) is considered to be a suitable choice for the DEMO primary heat transfer systems, being characterized by lower thermal inertia with respect to the most common U-tube steam generators. In this framework, the ENEA has undertaken construction of the STEAM facility at Brasimone R.C., aiming at characterizing the behavior of the DEMO OTSG and related water coolant systems in steady-state and transient conditions. A dedicated OTSG mock-up has been conceived and designed, adopting a scaling procedure, keeping the height 1:1 of the DEMO OTSGs. The conceptual design has been supported by RELAP5/Mod3.3 thermal-hydraulic calculations. CFD and FEM codes have been used for fluid-dynamic analyses and mechanical stress analyses, respectively, in specific parts of the component

Blind simulations of NACIE-UP experimental tests by STH codes

In the frame of the SESAME project, a benchmarking activity was proposed to validate the existing system thermal-hydraulics codes for Heavy Liquid Metal reactors. More specifically, blind simulations on three well-defined experiments were carried out on the NACIE-UP facility, using CATHARE by ENEA, ATHLET by GRS, RELAP5-3D by University of Roma and RELAP5/Mod3.3 by University of Pisa. The numerical models were calibrated in terms of system thermal losses and gas enhanced circulation by means of the outcomes from specific experimental preliminary tests. The present discussion expose, compare and analyze the numerical results of some representative parameters (primary lead-bismuth eutectic (LBE) mass flow rate, temperatures and pressure) charaterizing the system behaviour in transiet scenarios in a “pre-test” blind numerical assessment

STEAM Experimental Facility: A Step Forward for the Development of the EU DEMO BoP Water Coolant Technology

Within the EUROfusion roadmap for the technological development of the European-DEMOnstration (EU-DEMO) reactor, a key point has been identified in the discontinuous operation (pulse-dwell-pulse) of the machine. Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) adopt technology and components commonly used in nuclear fission power plants, whose performances could be negatively affected by the above mentioned pulsation, as well as by low-load operation in the dwell phase. This makes mandatory a full assessment of the functional feasibility of such components through accurate design and validation. For this purpose, ENEA Experimental Engineering Division at Brasimone R.C. aims at realizing STEAM, a water operated facility forming part of the multipurpose experimental infrastructure Water cooled lithium lead -thermal-HYDRAulic (W-HYDRA), conceived to investigate the water technologies applied to the DEMO BB and Balance of Plant systems and components. The experimental validation has the two main objectives of reproducing the DEMO operational phases by means of steady-state and transient tests, as well as performing dedicated tests on the steam generator aiming at demonstrating its ability to perform as intended during the power phases of the machine. STEAM is mainly composed of primary and secondary water systems reproducing the thermodynamic conditions of the DEMO WCLL BB PHTS and power conversion system, respectively. The significance of the STEAM facility resides in its capacity to amass experimental data relevant for the advancement of fusion-related technologies. This capability is attributable to the comprehensive array of instruments with which the facility will be equipped and whose strategic location is described in this work. The operational phases of the STEAM facility at different power levels are presented, according to the requirements of the experiments. Furthermore, a preliminary analysis for the definition of the control strategy for the OTSG mock-up was performed. In particular, two different control strategies were identified and tested, both keeping the primary mass flow constant and regulating the feedwater mass flow to follow a temperature set-point in the primary loop. The obtained numerical results yielded preliminary feedback on the regulation capability of the DEMO steam generator mock-up during pulsed operation, showing that no relevant overtemperature jeopardized the facility integrity, thanks to the high system responsivity to rapid load variations

Mock-ups fabrication by HRP technology with advanced W-alloy monoblocks for DEMO divertor target

Tungsten is the primary candidate armour material for the divertor target of the European demonstration fusion power plant. During operation at high temperature, pure tungsten is subject to fracture and recrystallization which results in a loss of strength and worsening of the thermal properties. Additionally, loss-of-coolant accidents with simultaneous air ingress can generate volatile and radioactive tungsten oxides. Advanced W-alloys were developed as alternative and upgrading armour materials of pure tungsten, such as potassium-doped tungsten laminates and self-passivating tungsten alloys. Three mock-ups were manufactured using potassium-doped tungsten laminates, W-10Cr-0.5Y and W-10Cr-0.5Y-0.5Zr as armour materials, each of them consisting of n degrees 4 blocks. The fabrication required optimization and upscaling of the ITER-like process which foresees oxygen-free high conductivity copper as interlayer joined to W-alloy armour block and CuCrZr ITER grade pipe welded to the Cu/W-alloy blocks by hot radial pressing. For quality control of the fabrication steps, non-destructive examination by ultrasonic testing was done on the monoblocks as received, after casting, after hot radial pressing and after high heat flux testing. The results demonstrated that these W-alloys can be used as armour materials of the European demonstration fusion power plant divertor target