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Dr. Fabio Giannetti
Sapienza University of Rome, DIAEE

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0 Nuclear Decommissioning
0 Nuclear Energy
0 Nuclear Engineering
0 Thermal Hydraulics

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Journal article
Published: 31 July 2021 in Nuclear Engineering and Design
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A MELCOR model of a boiling water reactor has been developed by Sapienza University of Rome in the framework of the MUSA project aiming at identifying and quantifying uncertainty sources in severe accidents analyses. To develop the model, the Fukushima Unit 3 boiling water reactor has been taken as reference, and a preliminary sensitivity analysis with RAVEN software coupled with MELCOR has been performed to quantify the influence of core degradation parameters on selected key figures-of-merit, such as pressure vessel and containment pressures, core liquid level, hydrogen generation, lower head breach time and source term. Results of base case satisfactorily predict main pressure and liquid level data measurement from TEPCO. As well as, results from the uncertainty analysis envelope the majority of thermohydraulic TEPCO reported measured data. Furthermore, median of calculated core status is in likely agreement with recent TEPCO containment inspections and muon measurements, with about 25% of fuel rods remained intact in RPV and 65% of core masses ejected to the pedestal.

ACS Style

Matteo D'Onorio; Alessio Giampaolo; Gianfranco Caruso; Fabio Giannetti. Preliminary uncertainty quantification of the core degradation models in predicting the Fukushima Daiichi unit 3 severe accident. Nuclear Engineering and Design 2021, 382, 111383 .

AMA Style

Matteo D'Onorio, Alessio Giampaolo, Gianfranco Caruso, Fabio Giannetti. Preliminary uncertainty quantification of the core degradation models in predicting the Fukushima Daiichi unit 3 severe accident. Nuclear Engineering and Design. 2021; 382 ():111383.

Chicago/Turabian Style

Matteo D'Onorio; Alessio Giampaolo; Gianfranco Caruso; Fabio Giannetti. 2021. "Preliminary uncertainty quantification of the core degradation models in predicting the Fukushima Daiichi unit 3 severe accident." Nuclear Engineering and Design 382, no. : 111383.

Journal article
Published: 03 June 2021 in Annals of Nuclear Energy
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Despite system thermal–hydraulic codes were extensively validated for transient simulations of LWR, several activities highlighted limited capabilities of these tools to model heat transfer within in-pool passive power removal system. Discrepancies with experimental results were related to the underestimate of pool boiling and film condensation heat transfer coefficients. Thus, the DIAEE of “Sapienza” University of Rome developed a modified version of RELAP5/Mod3.3, able to handle fundamental heat exchange phenomena involved in passive in-pool safety systems. A primary validation procedure has been performed for separated and integral effects. Dealing with nucleate boiling, the Root Mean Squared Relative Error (RMSRE) of wall superheat has been reduced from 1.290 to 0.182. Concerning film condensation, wall temperature RMSRE has been reduced from 0.192 to 0.058. The integral effect assessment has involved an experimental test of the PERSEO facility. The qualitative comparison between experiments and calculations has highlighted significant improvements of the modified RELAP5/Mod3.3.

ACS Style

Vincenzo Narcisi; Lorenzo Melchiorri; Fabio Giannetti. Improvements of RELAP5/Mod3.3 heat transfer capabilities for simulation of in-pool passive power removal systems. Annals of Nuclear Energy 2021, 160, 108436 .

AMA Style

Vincenzo Narcisi, Lorenzo Melchiorri, Fabio Giannetti. Improvements of RELAP5/Mod3.3 heat transfer capabilities for simulation of in-pool passive power removal systems. Annals of Nuclear Energy. 2021; 160 ():108436.

Chicago/Turabian Style

Vincenzo Narcisi; Lorenzo Melchiorri; Fabio Giannetti. 2021. "Improvements of RELAP5/Mod3.3 heat transfer capabilities for simulation of in-pool passive power removal systems." Annals of Nuclear Energy 160, no. : 108436.

Journal article
Published: 17 May 2021 in Fusion Engineering and Design
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EU-DEMO will be a DEMOnstration Fusion power plant designed to demonstrate production of grid electricity from fusion at the level of a few hundred MW. The Primary Heat Transfer System (PHTS) transfers heat from the breeding blanket (BB), divertor and vacuum vessel to the secondary Power Conversion System (PCS) responsible for conversion of thermal energy into electricity. Two main BB conceptions, and the relative PHTSs, for EU-DEMO are being developed: the Helium Cooled Pebble Bed (HCPB) BB and the Water Cooled Lithium Lead (WCLL) BB. Two options for each conception are considered: with or without the Intermediate Heat Transfer System (IHTS), containing the Energy Storage System (ESS), between the BB PHTS and PCS. The role of IHTS+ ESS is to ensure continuous smooth thermal energy transfer from the reactor sources to PCS despite the pulsed operation of the DEMO reactor. In the present work we discuss the mature concept of the PCS configuration for the option WCLL BB with the IHTS+ESS (based on the 2018 EU-DEMO reference), which allows almost constant production of electricity during both plasma pulse and dwell phases. The operating parameters of the circuit were optimized to minimize the temperature oscillations ΔT = |Tpulse - Tdwell| in all the circuit components, which occur due to the pulsation of the DEMO cold sources of Divertor and Vacuum Vessel whose HXs are integrated in PCS itself. Operation of the PCS circuit during the pulse and dwell phases was simulated using the GateCycle software, to show the system performance and to enable discussion on the feasibility of the concept.

ACS Style

Leszek Malinowski; Monika Lewandowska; Fabio Giannetti. Design and optimization of the secondary circuit for the WCLL BB option of the EU-DEMO power plant. Fusion Engineering and Design 2021, 169, 112642 .

AMA Style

Leszek Malinowski, Monika Lewandowska, Fabio Giannetti. Design and optimization of the secondary circuit for the WCLL BB option of the EU-DEMO power plant. Fusion Engineering and Design. 2021; 169 ():112642.

Chicago/Turabian Style

Leszek Malinowski; Monika Lewandowska; Fabio Giannetti. 2021. "Design and optimization of the secondary circuit for the WCLL BB option of the EU-DEMO power plant." Fusion Engineering and Design 169, no. : 112642.

Journal article
Published: 30 April 2021 in Fusion Engineering and Design
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In this paper, the conceptual design of the International Thermonuclear Experimental Reactor (ITER) Water-Cooled Lithium-Lead (WCLL) Test Blanket Module (TBM) Water Cooling System (WCS) from Europe is presented. The system consists of two loops in series. This design feature allows the removal of heat from the TBM box avoiding at the same time the release of radionuclides into the ITER Component Cooling Water System (CCWS), that acts as WCLL Test Blanket System heat sink. For this purpose, the WCS primary loop deals with the direct heat removal from the ITER TBM and the secondary one implements physical separation between the contaminated primary loop coolant and the CCWS. The insertion of an economizer into the primary loop determines the characteristic “eight” shape of the circuit. This choice was done in order to reduce the temperature difference on the intermediate heat exchanger. Hairpin type and steam bubble are the technologies selected for heat exchangers and pressurizers, respectively. Pressure and temperature control systems are foreseen to limit excursions from rated values in normal operational states and abnormal transients. A computational activity was promoted to assess the WCLL-WCS conceptual design, using a modified version of the RELAP5 Mod3.3 system code. A detailed thermal-hydraulic model was developed on the basis of design outcomes. The nodalization scheme includes the TBM, the WCS, a portion of the CCWS and the lithium-lead circuit. The computational campaign involved both the normal operational state and selected abnormal transients. In all the scenarios simulated, the conceptual design has highlighted the capability of operating the system respecting all the thermal-hydraulic requirements. The abnormal transient selected and presented is the loss of flow in the CCWS (loss of heat sink). In these conditions, TBM cooling function has been verified, keeping standard control strategies without any external action.

ACS Style

C. Ciurluini; V. Narcisi; A. Tincani; C. Ortiz Ferrer; F. Giannetti. Conceptual design overview of the ITER WCLL Water Cooling System and supporting thermal-hydraulic analysis. Fusion Engineering and Design 2021, 171, 112598 .

AMA Style

C. Ciurluini, V. Narcisi, A. Tincani, C. Ortiz Ferrer, F. Giannetti. Conceptual design overview of the ITER WCLL Water Cooling System and supporting thermal-hydraulic analysis. Fusion Engineering and Design. 2021; 171 ():112598.

Chicago/Turabian Style

C. Ciurluini; V. Narcisi; A. Tincani; C. Ortiz Ferrer; F. Giannetti. 2021. "Conceptual design overview of the ITER WCLL Water Cooling System and supporting thermal-hydraulic analysis." Fusion Engineering and Design 171, no. : 112598.

Journal article
Published: 23 March 2021 in Fusion Engineering and Design
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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.

ACS Style

M. Tarantino; P. Lorusso; A. Del Nevo; I. Di Piazza; F. Giannetti; D. Martelli. Preliminary design of a helical coil steam generator mock-up for the CIRCE facility for the development of DEMO LiPb heat exchanger. Fusion Engineering and Design 2021, 169, 112459 .

AMA Style

M. Tarantino, P. Lorusso, A. Del Nevo, I. Di Piazza, F. Giannetti, D. Martelli. Preliminary design of a helical coil steam generator mock-up for the CIRCE facility for the development of DEMO LiPb heat exchanger. Fusion Engineering and Design. 2021; 169 ():112459.

Chicago/Turabian Style

M. Tarantino; P. Lorusso; A. Del Nevo; I. Di Piazza; F. Giannetti; D. Martelli. 2021. "Preliminary design of a helical coil steam generator mock-up for the CIRCE facility for the development of DEMO LiPb heat exchanger." Fusion Engineering and Design 169, no. : 112459.

Journal article
Published: 19 March 2021 in Fusion Engineering and Design
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The European Research Roadmap to the Realisation of Fusion Energy foresees that the DEMO reactor is going to succeed ITER in the pathway towards the exploitation of nuclear fusion, achieving long plasma operation time, demonstrating tritium self-sufficiency and producing net electric output on an industrial scale. Therefore, its design must be more oriented towards the Balance of Plant (BoP) than it is in ITER. Since the early pre-conceptual phase of the DEMO project, emphasis has been laid on identifying the main requirements affecting the overall architecture of the BoP. For instance, specific efforts and proper solutions have been envisaged to cope with the pulsed nature of the heat source. Furthermore, the current development of two blanket concepts calls for two separate BoP options to be conceived. This paper summarizes the main alternatives outlined at the end of DEMO pre-conceptual design phase for the BoP concepts based on both the Helium-Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL) Breeding Blanket (BB) technologies. Then, the assumed reference configurations of both the BoP concepts are described in detail, highlighting the main features and the most relevant engineering aspects. Attention will be focussed on technological challenges, integration constraints and other open issues, highlighting pros and cons of the chosen BoP options to be further investigated in the next design phase.

ACS Style

L. Barucca; E. Bubelis; S. Ciattaglia; A. D’Alessandro; A. Del Nevo; F. Giannetti; W. Hering; P. Lorusso; E. Martelli; I. Moscato; A. Quartararo; A. Tarallo; E. Vallone. Pre-conceptual design of EU DEMO balance of plant systems: Objectives and challenges. Fusion Engineering and Design 2021, 169, 112504 .

AMA Style

L. Barucca, E. Bubelis, S. Ciattaglia, A. D’Alessandro, A. Del Nevo, F. Giannetti, W. Hering, P. Lorusso, E. Martelli, I. Moscato, A. Quartararo, A. Tarallo, E. Vallone. Pre-conceptual design of EU DEMO balance of plant systems: Objectives and challenges. Fusion Engineering and Design. 2021; 169 ():112504.

Chicago/Turabian Style

L. Barucca; E. Bubelis; S. Ciattaglia; A. D’Alessandro; A. Del Nevo; F. Giannetti; W. Hering; P. Lorusso; E. Martelli; I. Moscato; A. Quartararo; A. Tarallo; E. Vallone. 2021. "Pre-conceptual design of EU DEMO balance of plant systems: Objectives and challenges." Fusion Engineering and Design 169, no. : 112504.

Journal article
Published: 16 March 2021 in Fusion Engineering and Design
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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.

ACS Style

P. Lorusso; E. Martelli; A. Del Nevo; V. Narcisi; F. Giannetti; M. Tarantino. Development of a PbLi heat exchanger for EU DEMO fusion reactor: Experimental test and system code assessment. Fusion Engineering and Design 2021, 169, 112462 .

AMA Style

P. Lorusso, E. Martelli, A. Del Nevo, V. Narcisi, F. Giannetti, M. Tarantino. Development of a PbLi heat exchanger for EU DEMO fusion reactor: Experimental test and system code assessment. Fusion Engineering and Design. 2021; 169 ():112462.

Chicago/Turabian Style

P. Lorusso; E. Martelli; A. Del Nevo; V. Narcisi; F. Giannetti; M. Tarantino. 2021. "Development of a PbLi heat exchanger for EU DEMO fusion reactor: Experimental test and system code assessment." Fusion Engineering and Design 169, no. : 112462.

Journal article
Published: 11 March 2021 in Energies
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The Breeding Blanket (BB) is one of the key components of the European Demonstration (EU-DEMO) fusion reactor. Its main subsystems, the Breeder Zone (BZ) and the First Wall (FW), are cooled by two independent cooling circuits, called Primary Heat Transfer Systems (PHTS). Evaluating the BB PHTS performances in anticipated transient and accident conditions is a relevant issue for the design of these cooling systems. Within the framework of the EUROfusion Work Package Breeding Blanket, it was performed a thermal-hydraulic analysis of the PHTS during transient conditions belonging to the category of “Decrease in Coolant System Flow Rate”, by using Reactor Excursion Leak Analysis Program (RELAP5) Mod3.3. The BB, the PHTS circuits, the BZ Once Through Steam Generators and the FW Heat Exchangers were included in the study. Selected transients consist in partial and complete Loss of Flow Accident (LOFA) involving either the BZ or the FW PHTS Main Coolant Pumps (MCPs). The influence of the loss of off-site power, combined with the accident occurrence, was also investigated. The transient analysis was performed with the aim of design improvement. The current practice of a standard Pressurized Water Reactor (PWR) was adopted to propose and study actuation logics related to each accidental scenario. The appropriateness of the current PHTS design was demonstrated by simulation outcomes.

ACS Style

Cristiano Ciurluini; Fabio Giannetti; Alessandro Del Nevo; Gianfranco Caruso. Study of the EU-DEMO WCLL Breeding Blanket Primary Cooling Circuits Thermal-Hydraulic Performances during Transients Belonging to LOFA Category. Energies 2021, 14, 1541 .

AMA Style

Cristiano Ciurluini, Fabio Giannetti, Alessandro Del Nevo, Gianfranco Caruso. Study of the EU-DEMO WCLL Breeding Blanket Primary Cooling Circuits Thermal-Hydraulic Performances during Transients Belonging to LOFA Category. Energies. 2021; 14 (6):1541.

Chicago/Turabian Style

Cristiano Ciurluini; Fabio Giannetti; Alessandro Del Nevo; Gianfranco Caruso. 2021. "Study of the EU-DEMO WCLL Breeding Blanket Primary Cooling Circuits Thermal-Hydraulic Performances during Transients Belonging to LOFA Category." Energies 14, no. 6: 1541.

Journal article
Published: 27 February 2021 in Nuclear Engineering and Design
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In the framework of the HORIZON2020 SESAME European project, an experimental campaign has been carried out on the large Lead-Bismuth Eutectic pool integral effect CIRCE facility at ENEA Brasimone Research Center, implementing the HERO test section. This test section is aimed at supporting the development of the ALFRED design. Within the SESAME Project, three protected loss of flow accident tests have been designed and executed, identified as SE-Test1, SE-Test2, and SE-Test3. The third test (SE-Test3) has been selected by the project participants for a benchmark activity involving system thermal-hydraulic codes and coupled system thermalhydraulic/ CFD. This activity has been divided into two phases: a blind phase, and a post-test phase. The present paper illustrates the results achieved during the post-test phase of the CIRCE-HERO benchmark activity. Four participants simulated the experimental test. Three (ENEA, SCKCEN and UNIROMA1) used different versions of RELAP5 code, i.e. RELAP5-3D and RELAP5Mod3.3, and one (NRG) used a coupled approach based on an in-house system thermal-hydraulic code (SPECTRA) and the commercial code ANSYS CFX. The benchmark activity hereafter presented is a relevant exercise for evaluating and comparing the predictive capabilities of system thermal-hydraulic codes, CFD codes and coupled technics, in relation to phenomena occurring during the transition between forced and natural circulation (e.g. loss of flow) in a heavy liquid metals Generation IV system. The results of the numerical exercise showed an adequate capability of the codes to reproduce the relevant phenomena involved during the experiment. The SYS-TH codes have been more accurate than SYS-TH/CFD in predicting the trend of the main parameters during the transient, reproducing them with quite satisfactory accuracy. SYS-TH/CFD simulation, instead, provided a more satisfactory representation of the pool thermal stratification. The results highlighted that the planning and the execution of experiments fully devoted for the code V&V process is needed for the further development of such numerical tools, in particular for SYS-TH/CFD coupled approaches, whose use is more recent respect to SYS-TH codes.

ACS Style

P. Lorusso; A. Del Nevo; V. Narcisi; F. Giannetti; G. Caruso; K. Zwijsen; P.A. Breijder; T. Hamidouche; D. Castelliti; D. Rozzia; M. Tarantino. Total loss of flow benchmark in CIRCE-HERO integral test facility. Nuclear Engineering and Design 2021, 376, 111086 .

AMA Style

P. Lorusso, A. Del Nevo, V. Narcisi, F. Giannetti, G. Caruso, K. Zwijsen, P.A. Breijder, T. Hamidouche, D. Castelliti, D. Rozzia, M. Tarantino. Total loss of flow benchmark in CIRCE-HERO integral test facility. Nuclear Engineering and Design. 2021; 376 ():111086.

Chicago/Turabian Style

P. Lorusso; A. Del Nevo; V. Narcisi; F. Giannetti; G. Caruso; K. Zwijsen; P.A. Breijder; T. Hamidouche; D. Castelliti; D. Rozzia; M. Tarantino. 2021. "Total loss of flow benchmark in CIRCE-HERO integral test facility." Nuclear Engineering and Design 376, no. : 111086.

Journal article
Published: 25 February 2021 in Fusion Engineering and Design
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The Water Cooled Lithium Lead Test Blanket System (WCLL TBS) is one of the EU Test Blanket Systems candidate for being installed and operated in ITER. In view of its Conceptual Design Review by F4E and ITER Organization (IO), planned for mid-September 2020, several technical activities have been performed in the areas of WCLL TBS Ancillary Systems design. In this article the outcomes of the conceptual design phase of the four main Ancillary Systems of WCLL TBS, namely the Water Cooling System (WCS), the Coolant Purification System (CPS), the PbLi loop and the Tritium Extraction System (TES), are reported and critically discussed. In particular, for each Ancillary System hereafter are reported: i) a short design description, including the conceptual design of their main components together with their operative conditions under the so-called Normal Operational State (NOS), ii) the ESP-ESPN classification for their main components, and iii) their arrangement and integration in the assigned ITER areas (PC#16, Vertical Shaft, TCWS Vault, Galleries and Tritium Process Room).

ACS Style

Amelia Tincani; Pietro Arena; Maurizio Bruzzone; Ilenia Catanzaro; Cristiano Ciurluini; Alessandro Del Nevo; Pietro Alessandro Di Maio; Ruggero Forte; Fabio Giannetti; Stefano Lorenzi; Emanuela Martelli; Carlos Moreno; Rocco Mozzillo; Carlos Ortiz; Ferruccio Paoletti; Veronica Pierantoni; Italo Ricapito; Gandolfo Alessandro Spagnuolo; Andrea Tarallo; Claudio Tripodo; Antonio Cammi; Marco Utili; Konstantina Voukelatou; Erik Walcz; Balazs Lesko; Jessica Korzeniowska; Pierluigi Chiovaro; Vincenzo Narcisi. Conceptual design of the main Ancillary Systems of the ITER Water Cooled Lithium Lead Test Blanket System. Fusion Engineering and Design 2021, 167, 112345 .

AMA Style

Amelia Tincani, Pietro Arena, Maurizio Bruzzone, Ilenia Catanzaro, Cristiano Ciurluini, Alessandro Del Nevo, Pietro Alessandro Di Maio, Ruggero Forte, Fabio Giannetti, Stefano Lorenzi, Emanuela Martelli, Carlos Moreno, Rocco Mozzillo, Carlos Ortiz, Ferruccio Paoletti, Veronica Pierantoni, Italo Ricapito, Gandolfo Alessandro Spagnuolo, Andrea Tarallo, Claudio Tripodo, Antonio Cammi, Marco Utili, Konstantina Voukelatou, Erik Walcz, Balazs Lesko, Jessica Korzeniowska, Pierluigi Chiovaro, Vincenzo Narcisi. Conceptual design of the main Ancillary Systems of the ITER Water Cooled Lithium Lead Test Blanket System. Fusion Engineering and Design. 2021; 167 ():112345.

Chicago/Turabian Style

Amelia Tincani; Pietro Arena; Maurizio Bruzzone; Ilenia Catanzaro; Cristiano Ciurluini; Alessandro Del Nevo; Pietro Alessandro Di Maio; Ruggero Forte; Fabio Giannetti; Stefano Lorenzi; Emanuela Martelli; Carlos Moreno; Rocco Mozzillo; Carlos Ortiz; Ferruccio Paoletti; Veronica Pierantoni; Italo Ricapito; Gandolfo Alessandro Spagnuolo; Andrea Tarallo; Claudio Tripodo; Antonio Cammi; Marco Utili; Konstantina Voukelatou; Erik Walcz; Balazs Lesko; Jessica Korzeniowska; Pierluigi Chiovaro; Vincenzo Narcisi. 2021. "Conceptual design of the main Ancillary Systems of the ITER Water Cooled Lithium Lead Test Blanket System." Fusion Engineering and Design 167, no. : 112345.

Journal article
Published: 26 January 2021 in Energies
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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.

ACS Style

Pierdomenico Lorusso; Emanuela Martelli; Alessandro Del Nevo; Vincenzo Narcisi; Fabio Giannetti; Mariano Tarantino. Experimental Investigation on CIRCE-HERO for the EU DEMO PbLi/Water Heat Exchanger Development. Energies 2021, 14, 628 .

AMA Style

Pierdomenico Lorusso, Emanuela Martelli, Alessandro Del Nevo, Vincenzo Narcisi, Fabio Giannetti, Mariano Tarantino. Experimental Investigation on CIRCE-HERO for the EU DEMO PbLi/Water Heat Exchanger Development. Energies. 2021; 14 (3):628.

Chicago/Turabian Style

Pierdomenico Lorusso; Emanuela Martelli; Alessandro Del Nevo; Vincenzo Narcisi; Fabio Giannetti; Mariano Tarantino. 2021. "Experimental Investigation on CIRCE-HERO for the EU DEMO PbLi/Water Heat Exchanger Development." Energies 14, no. 3: 628.

Journal article
Published: 20 January 2021 in Fusion Engineering and Design
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The water-cooled EU-DEMO Breeding Blanket (BB) is cooled by two independent circuits, the Breeder Zone (BZ) and the First Wall (FW) Primary Heat Transfer Systems (PHTS). The configuration under study foresees the presence of an Intermediate Heat Transfer System and an Energy Storage System to operate the turbine during both the pulse (2 h) and the dwell time (10 min) at almost constant load, despite the plasma power pulsation. Within the framework of the EUROfusion WPBOP research activity, a RELAP5/Mod3.3 model was developed to investigate the thermal-hydraulic behavior of the primary cooling systems during transient conditions belonging to the category of “Decrease in Coolant System Flow Rate”. The nodalization includes the BB, the PHTS circuits, the BZ Once Through Steam Generators and the FW Heat EXchangers. The model was initially used to simulate the nominal conditions with the pulse and dwell phases. Then, starting from the pulse, a Loss of Flow Accident (LOFA) was selected to preliminary evaluate the PHTS behavior with the aim of the design improvement. LOFA analyses were performed considering the complete loss of both the FW and BZ PHTS main coolant pumps (MCPs). A sensitivity was carried out to assess the impact of the MCPs flywheel on the main PHTS parameters. Transient results highlighted the appropriateness of the current design with no need for further mitigation actions.

ACS Style

Cristiano Ciurluini; Fabio Giannetti; Emanuela Martelli; Alessandro Del Nevo; Luciana Barucca; Gianfranco Caruso. Analysis of the thermal-hydraulic behavior of the EU-DEMO WCLL breeding blanket cooling systems during a loss of flow accident. Fusion Engineering and Design 2021, 164, 112206 .

AMA Style

Cristiano Ciurluini, Fabio Giannetti, Emanuela Martelli, Alessandro Del Nevo, Luciana Barucca, Gianfranco Caruso. Analysis of the thermal-hydraulic behavior of the EU-DEMO WCLL breeding blanket cooling systems during a loss of flow accident. Fusion Engineering and Design. 2021; 164 ():112206.

Chicago/Turabian Style

Cristiano Ciurluini; Fabio Giannetti; Emanuela Martelli; Alessandro Del Nevo; Luciana Barucca; Gianfranco Caruso. 2021. "Analysis of the thermal-hydraulic behavior of the EU-DEMO WCLL breeding blanket cooling systems during a loss of flow accident." Fusion Engineering and Design 164, no. : 112206.

Journal article
Published: 19 October 2020 in Fusion Engineering and Design
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The Water-Cooled Lithium-Lead Breeding Blanket is a key component of a fusion power plant, in charge of ensure Tritium production, shield Vacuum Vessel and magnets and remove the heat power deposited by radiation and particles arising from plasma. The last function is fulfilled by First Wall and Breeding Zone independent cooling systems. Several layouts of BZ coolant system have been investigated in the last years to identify a configuration that might guarantee EUROFER temperature below the limit (550 °C) and good thermal-hydraulic performances (i.e. water outlet temperature of 328 °C). A research activity is conducted to study and compare different modelling approaches to simulate the heat transfer within the BZ liquid metal, assessing their impact on the numerical prediction of the WCLL blanket thermal performances. An approach will rely on the simulation of convective and diffusive heat transfer processes taking place within the liquid metal by means of a CFD tool based on the Finite Volume Method. Conversely, the other approach will roughly assume a pure diffusive heat transfer mechanism within the breeder, due to the very low velocities envisaged for its flow field. In this case the heat transfer performances will be preferably assessed by means of a commercial code based on the Finite Element Method. The analyses have been carried out with reference to the so called “WCLL BB 2018 V0.6″ equatorial cell. Advantages and issues from the thermal-hydraulic point of view are identified, the impact of the imposed boundary conditions and heat transfer properties, with the implemented correlations, on the respective results is critically discussed.

ACS Style

Francesco Edemetti; Emanuela Martelli; Alessandro Del Nevo; Fabio Giannetti; Pietro Arena; Ruggero Forte; Pietro Alessandro Di Maio; Gianfranco Caruso. On the impact of the heat transfer modelling approach on the prediction of EU-DEMO WCLL breeding blanket thermal performances. Fusion Engineering and Design 2020, 161, 112051 .

AMA Style

Francesco Edemetti, Emanuela Martelli, Alessandro Del Nevo, Fabio Giannetti, Pietro Arena, Ruggero Forte, Pietro Alessandro Di Maio, Gianfranco Caruso. On the impact of the heat transfer modelling approach on the prediction of EU-DEMO WCLL breeding blanket thermal performances. Fusion Engineering and Design. 2020; 161 ():112051.

Chicago/Turabian Style

Francesco Edemetti; Emanuela Martelli; Alessandro Del Nevo; Fabio Giannetti; Pietro Arena; Ruggero Forte; Pietro Alessandro Di Maio; Gianfranco Caruso. 2020. "On the impact of the heat transfer modelling approach on the prediction of EU-DEMO WCLL breeding blanket thermal performances." Fusion Engineering and Design 161, no. : 112051.

Conference paper
Published: 01 August 2020 in Journal of Physics: Conference Series
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Cristiano Ciurluini; Vincenzo Narcisi; Fabio Giannetti; Luca Cretara; Gianfranco Caruso. Preliminary neutron kinetic – thermal hydraulic coupled analysis of the ALFRED reactor using PHISICS/RELAP5-3D. Journal of Physics: Conference Series 2020, 1599, 1 .

AMA Style

Cristiano Ciurluini, Vincenzo Narcisi, Fabio Giannetti, Luca Cretara, Gianfranco Caruso. Preliminary neutron kinetic – thermal hydraulic coupled analysis of the ALFRED reactor using PHISICS/RELAP5-3D. Journal of Physics: Conference Series. 2020; 1599 ():1.

Chicago/Turabian Style

Cristiano Ciurluini; Vincenzo Narcisi; Fabio Giannetti; Luca Cretara; Gianfranco Caruso. 2020. "Preliminary neutron kinetic – thermal hydraulic coupled analysis of the ALFRED reactor using PHISICS/RELAP5-3D." Journal of Physics: Conference Series 1599, no. : 1.

Journal article
Published: 30 June 2020 in Nuclear Technology
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In the framework of the European Union MAXSIMA project, the safety of the steam generator (SG) adopted in the primary loop of the Heavy Liquid Metal Fast Reactor has been studied investigating the consequences and damage propagation of a SG tube rupture event and characterizing leak rates from typical cracks. Instrumentation able to promptly detect the presence of a crack in the SG tubes may be used to prevent its further propagation, which would lead to a full rupture of the tube. Application of the leak-before-break concept is relevant for improving the safety of a reactor system and decreasing the probability of a pipe break event. In this framework, a new experimental campaign (Test Series C) has been carried out in the LIFUS5/Mod3 facility, installed at ENEA Centro Ricerche Brasimone, in order to characterize and to correlate the leak rate through typical cracks occurring in the pressurized tubes with signals detected by proper transducers. Test C1.3_60 was executed injecting water at about 20 bars and 200°C into lead-bismuth eutectic alloy. The injection was performed through a laser microholed plate 60 μm in diameter. Analysis of the thermohydraulic data permitted characterization of the leakage through typical cracks that can occur in the pressurized tubes of the SG. Analysis of the data acquired by microphones and accelerometers highlighted that it is possible to correlate the signals to the leakage and the rate of release.

ACS Style

Marica Eboli; Alessandro Del Nevo; Nicola Forgione; Fabio Giannetti; Daniele Mazzi; Marco Ramacciotti. Experimental Characterization of Leak Detection Systems in HLM Pool Using LIFUS5/Mod3 Facility. Nuclear Technology 2020, 206, 1409 -1420.

AMA Style

Marica Eboli, Alessandro Del Nevo, Nicola Forgione, Fabio Giannetti, Daniele Mazzi, Marco Ramacciotti. Experimental Characterization of Leak Detection Systems in HLM Pool Using LIFUS5/Mod3 Facility. Nuclear Technology. 2020; 206 (9):1409-1420.

Chicago/Turabian Style

Marica Eboli; Alessandro Del Nevo; Nicola Forgione; Fabio Giannetti; Daniele Mazzi; Marco Ramacciotti. 2020. "Experimental Characterization of Leak Detection Systems in HLM Pool Using LIFUS5/Mod3 Facility." Nuclear Technology 206, no. 9: 1409-1420.

Journal article
Published: 07 June 2020 in Fusion Engineering and Design
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In the frame of the EUROfusion roadmap for the development of the DEMO power plant, a research activity was carried out to develop a Lithium–Lead/water heat exchanger. The component should be capable to remove nuclear heat deposited in the liquid metal of the Dual Coolant Lithium Lead breeding blanket and feeding a steam turbine, ensuring an efficient thermal power conversion to electricity. One of the selected configurations is the steam generator bayonet tube. The HERO test section is an experimental mock-up in a relevant scale of this steam generator, consisting of a bundle of seven double-wall bayonet tubes with a leakage monitoring system. This test section, developed by ENEA at Brasimone Research Center and installed in the main vessel of the CIRCE pool facility, aims to investigate the thermal-hydraulic features of the system, providing a database for thermal-hydraulic system codes validation. An experimental campaign was carried out to demonstrate technological feasibility and performances of the prototypical heat exchanger, suitable as steam generator for the PbLi loop of the Dual Coolant Lithium Lead and Water Cooled Lithium Lead breeding blankets. A post-test analysis has been realized with RELAP5-3D and RELAP5/Mod3.3 codes in order to evaluate code capability in simulating heat transfer in liquid metal side.

ACS Style

E. Martelli; A. Del Nevo; P. Lorusso; F. Giannetti; V. Narcisi; M. Tarantino. Investigation of heat transfer in a steam generator bayonet tube for the development of PbLi technology for EU DEMO fusion reactor. Fusion Engineering and Design 2020, 159, 111772 .

AMA Style

E. Martelli, A. Del Nevo, P. Lorusso, F. Giannetti, V. Narcisi, M. Tarantino. Investigation of heat transfer in a steam generator bayonet tube for the development of PbLi technology for EU DEMO fusion reactor. Fusion Engineering and Design. 2020; 159 ():111772.

Chicago/Turabian Style

E. Martelli; A. Del Nevo; P. Lorusso; F. Giannetti; V. Narcisi; M. Tarantino. 2020. "Investigation of heat transfer in a steam generator bayonet tube for the development of PbLi technology for EU DEMO fusion reactor." Fusion Engineering and Design 159, no. : 111772.

Journal article
Published: 16 May 2020 in Fusion Engineering and Design
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In this early development phase of the DEMO design the uncertainty affecting many operational and design parameters can modify main outcomes of accident scenario aiming at studying the critical conditions for the vacuum vessel and the contiguous containment volumes. The aim of this paper is to perform a preliminary sensitivity analysis of an accident progression predicted by MELCOR code considering selected parameters as a figure of merit to predict possible code outcomes. The uncertainty band will be evaluated through sensitivity analyses programmed, collected and statistically manipulated through RAVEN software tool. MELCOR and RAVEN have been internally coupled through a new Python code interface developed by Sapienza University of Rome, to perform sensitivity and uncertainty quantification analyses during severe accident transient. The Beyond Design Basis Accident (BDBA) analysis of an ex-vessel loss of coolant accident (LOCA) for the water-cooled lithium lead (WCLL) blanket concept has been simulated with the fusion version of MELCOR 1.8.6 code. The postulated initiating event (PIE) is a double-ended break in the first wall (FW) cooling system distributor ring, with simultaneous failure of the plasma shutdown system. An in-vessel breach of the coolant system occurs because of FW failure, with consequent unmitigated plasma shutdown transient. Sensitivity analysis results have shown that the FW temperature at which plasma in-vessel breach occurs is strongly correlated with the mass of hydrogen produced. The same parameter has also an impact on the overall accident scenario, such as the trigger of VVPSS rupture disks and thus source term mobilization.

ACS Style

Matteo D’Onorio; Fabio Giannetti; Maria Teresa Porfiri; Gianfranco Caruso. Preliminary sensitivity analysis for an ex-vessel LOCA without plasma shutdown for the EU DEMO WCLL blanket concept. Fusion Engineering and Design 2020, 158, 111745 .

AMA Style

Matteo D’Onorio, Fabio Giannetti, Maria Teresa Porfiri, Gianfranco Caruso. Preliminary sensitivity analysis for an ex-vessel LOCA without plasma shutdown for the EU DEMO WCLL blanket concept. Fusion Engineering and Design. 2020; 158 ():111745.

Chicago/Turabian Style

Matteo D’Onorio; Fabio Giannetti; Maria Teresa Porfiri; Gianfranco Caruso. 2020. "Preliminary sensitivity analysis for an ex-vessel LOCA without plasma shutdown for the EU DEMO WCLL blanket concept." Fusion Engineering and Design 158, no. : 111745.

Journal article
Published: 23 April 2020 in Fusion Engineering and Design
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The Water Cooled Lithium Lead (WCLL) is one of the selected breeding blanket (BB) concepts to be investigated in the EUROfusion Breeding Blanket Project (WPBB), and it was also recently chosen as one of the mock-up for ITER Test Blanket Module (TBM) program. The program foresees the test of different BB mock-ups, called Test Blanket Modules, with all the related ancillary systems. A pre-conceptual design of the Water Cooling System (WCS) of the ITER WCLL-TBM was developed considering the same cooling function of the EU-DEMO WCLL-BB primary heat transfer system (PHTS), but matching different boundary conditions: a scaled source power and far lower heat sink temperatures. A complete thermal-hydraulic (TH) model of the WCS loop and TBM set was developed using a modified version of RELAP5/Mod3.3 system code to verify component sizing and to investigate the system behavior during steady-state and transient conditions. The full plasma power scenario was simulated and used as an initial condition for transient calculations. ITER Normal Operational State (NOS) was studied to evaluate the system response. Simulation results highlighted the need for an electric heater to keep the WCS system in stable operation. A sensitivity analysis was carried out to optimize the heater duty cycle.

ACS Style

Cristiano Ciurluini; Fabio Giannetti; Amelia Tincani; Alessandro Del Nevo; Gianfranco Caruso; Italo Ricapito; Fabio Cismondi. Thermal-hydraulic modeling and analysis of the Water Cooling System for the ITER Test Blanket Module. Fusion Engineering and Design 2020, 158, 111709 .

AMA Style

Cristiano Ciurluini, Fabio Giannetti, Amelia Tincani, Alessandro Del Nevo, Gianfranco Caruso, Italo Ricapito, Fabio Cismondi. Thermal-hydraulic modeling and analysis of the Water Cooling System for the ITER Test Blanket Module. Fusion Engineering and Design. 2020; 158 ():111709.

Chicago/Turabian Style

Cristiano Ciurluini; Fabio Giannetti; Amelia Tincani; Alessandro Del Nevo; Gianfranco Caruso; Italo Ricapito; Fabio Cismondi. 2020. "Thermal-hydraulic modeling and analysis of the Water Cooling System for the ITER Test Blanket Module." Fusion Engineering and Design 158, no. : 111709.

Journal article
Published: 22 April 2020 in Nuclear Engineering and Design
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The Advanced Lead Fast Reactor European Demonstrator (ALFRED) was conceived in the framework of the LEADER (Lead-cooled European Advanced DEmonstration Reactor) project since 2010. Recently, a revised concept of the Reactor Coolant System (RCS) is ongoing to solve some identified issues such as thermal stratification and others. To verify the improvements of the ALFRED design the authors developed a detailed nodalization scheme of the reactor using RELAP5-3D© code. This paper provides a summary of the main improvements introduced in the revised concept of ALFRED reactor and a detailed discussion of the thermal-hydraulic analysis performed with RELAP5-3D©. The nodalization scheme is described in detail together with transient boundary conditions and timing of events. The numerical activity demonstrates the improvements of the revised configuration of the reactor that avoid the establishment of relevant thermal stratification phenomena in both normal and accidental operations. The effect of two relevant parameters is also investigated, highlighting that safety conditions are maintained in the whole spectrum of the sensitivity analysis.

ACS Style

Vincenzo Narcisi; Fabio Giannetti; Marco Caramello; Gianfranco Caruso. Preliminary evaluation of ALFRED revised concept under station blackout. Nuclear Engineering and Design 2020, 364, 110648 .

AMA Style

Vincenzo Narcisi, Fabio Giannetti, Marco Caramello, Gianfranco Caruso. Preliminary evaluation of ALFRED revised concept under station blackout. Nuclear Engineering and Design. 2020; 364 ():110648.

Chicago/Turabian Style

Vincenzo Narcisi; Fabio Giannetti; Marco Caramello; Gianfranco Caruso. 2020. "Preliminary evaluation of ALFRED revised concept under station blackout." Nuclear Engineering and Design 364, no. : 110648.

Journal article
Published: 21 February 2020 in Fusion Engineering and Design
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In-vessel Loss Of Coolant Accident (LOCA) is one of the Design Basis Accident to be considered to support the future DEMOnstration power plant safety assessment. The water-cooled lithium-lead (WCLL) Breeding Blanket (BB) concept relies on Lithium-Lead as breeder, neutron multiplier and tritium carrier. The breeding modules are cooled by two independent pressurized water systems: the fist-wall (FW) and the breeding zone (BZ) coolant systems. The postulated initiating event (PIE) considered for this safety analysis is a double ended pipe rupture of the blanket module first wall channels. This event causes the inlet of coolant into the plasma chamber volume triggering an unmitigated plasma disruption and the pressurization of the Vacuum Vessel (VV) volume. The fusion version of MELCOR code (ver. 1.8.6) is used to evaluate accident consequences for two different scenarios, with the presence and absence of the downstream isolation valves, respectively. The chemical reaction between the coolant and the first wall tungsten layer inside the VV has been considered together with the mobilization of the radioactive source term. Pressure and temperature transient behavior in the tokamak volumes demonstrate that safety margins are respected during the accidental sequence.

ACS Style

Matteo D’Onorio; Fabio Giannetti; Maria Teresa Porfiri; Gianfranco Caruso. Preliminary safety analysis of an in-vessel LOCA for the EU-DEMO WCLL blanket concept. Fusion Engineering and Design 2020, 155, 111560 .

AMA Style

Matteo D’Onorio, Fabio Giannetti, Maria Teresa Porfiri, Gianfranco Caruso. Preliminary safety analysis of an in-vessel LOCA for the EU-DEMO WCLL blanket concept. Fusion Engineering and Design. 2020; 155 ():111560.

Chicago/Turabian Style

Matteo D’Onorio; Fabio Giannetti; Maria Teresa Porfiri; Gianfranco Caruso. 2020. "Preliminary safety analysis of an in-vessel LOCA for the EU-DEMO WCLL blanket concept." Fusion Engineering and Design 155, no. : 111560.