Energies

Research

26 pages, 6833 KiB

Open AccessArticle

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

by Alessandra Vannoni, Pierdomenico Lorusso, Pietro Arena, Marica Eboli, Ranieri Marinari, Amelia Tincani, Cristiano Ciurluini, Fabio Giannetti, Nicolò Badodi, Claudio Tripodo, Antonio Cammi, Luciana Barucca, Andrea Tarallo, Pietro Agostini and Alessandro Del Nevo

Energies 2023, 16(23), 7811; https://doi.org/10.3390/en16237811 - 27 Nov 2023

Cited by 3 | Viewed by 678

Abstract

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 [...] Read more.

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. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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21 pages, 12254 KiB

Open AccessArticle

The Design of Water Loop Facility for Supporting the WCLL Breeding Blanket Technology and Safety

by Alessandra Vannoni, Pietro Arena, Bruno Gonfiotti, Marica Eboli, Pierdomenico Lorusso, Amelia Tincani, Nicolò Badodi, Antonio Cammi, Fabio Giannetti, Cristiano Ciurluini, Nicola Forgione, Francesco Galleni, Ilenia Catanzaro, Eugenio Vallone, Pietro Alessandro Di Maio, Pietro Agostini and Alessandro Del Nevo

Energies 2023, 16(23), 7746; https://doi.org/10.3390/en16237746 - 24 Nov 2023

Cited by 4 | Viewed by 705

Abstract

The WCLL Breeding Blanket of DEMO and the Test Blanket Module (TBM) of ITER require accurate R&D activities, i.e., concept validation at a relevant scale and safety demonstrations. In view of this, the strategic objective of the Water Loop (WL) facility, belonging to [...] Read more.

The WCLL Breeding Blanket of DEMO and the Test Blanket Module (TBM) of ITER require accurate R&D activities, i.e., concept validation at a relevant scale and safety demonstrations. In view of this, the strategic objective of the Water Loop (WL) facility, belonging to the W-HYDRA experimental platform planned at C.R. Brasimone of ENEA, is twofold: to conduct R&D activities for the WCLL BB to validate design performances and to increase the technical maturity level for selection and validation phases, as well as to support the ITER WCLL Test Blanket System program. Basically, the Water Loop facility will have the capability to investigate the design features and performances of scaled-down or portions of breeding blanket components, as well as full-scale TBM mock-ups. It is a large-/medium-scale water coolant plant that will provide water coolant at high pressure and temperature. It is composed by single-phase primary (designed at 18.5 MPa and 350 °C) and secondary (designed at 2.5 MPa and 220 °C) systems thermally connected with a two-phase tertiary loop acting as an ultimate heat sink (designed at 6 bar and 80 °C). The primary loop has two main sources of power: an electrical heater up to about 1 MWe, installed in the cold side, downstream of the pump and upstream of the test section, and an electron beam gun acting as a heat flux generator. The WL has unique features and is designed as a multi-purpose facility capable of being coupled with the LIFUS5/Mod4 facility to study PbLi/water reaction at a large scale. This paper presents the status of the Water Loop facility, highlighting objectives, design features, and the analyses performed. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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25 pages, 5669 KiB

Open AccessArticle

PbLi/Water Reaction: Experimental Campaign and Modeling Advancements in WPBB EUROfusion Project

by Marica Eboli, Pietro Arena, Nicolò Badodi, Antonio Cammi, Cristiano Ciurluini, Vittorio Cossu, Nicola Forgione, Francesco Galleni, Fabio Giannetti, Bruno Gonfiotti, Daniele Martelli, Lorenzo Melchiorri, Carmine Risi, Alessandro Tassone and Alessandro Del Nevo

Energies 2023, 16(23), 7729; https://doi.org/10.3390/en16237729 - 23 Nov 2023

Cited by 1 | Viewed by 713

Abstract

The Water-Cooled Lithium–Lead blanket concept is a candidate breeding blanket concept for the EU DEMO reactor and it is going to be tested as one of the Test Blanket Modules (TBM) inside the ITER reactor. A major safety issue for its design is [...] Read more.

The Water-Cooled Lithium–Lead blanket concept is a candidate breeding blanket concept for the EU DEMO reactor and it is going to be tested as one of the Test Blanket Modules (TBM) inside the ITER reactor. A major safety issue for its design is the interaction between PbLi and water caused by a tube rupture in the breeding zone, the so-called in-box LOCA (Loss of Coolant Accident) scenario. This issue has been investigated in the framework of FP8 EUROfusion Project Horizon 2020 and is currently ongoing in FP9 EUROfusion Horizon Europe, defining a strategy for addressing and solving WCLL in-box LOCA. This paper discusses the efforts pursued in recent years to deal with this key safety issue, providing a general view of the approach, a timeline, research and development, and experimental activities. These are conducted to master dominant phenomena and processes relevant to safety aspects during the postulated accident, to enhance the predictive capability and reliability of selected numerical tools, and to validate and qualify models and codes and the procedures for their applications, including coupling and chains of codes. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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10 pages, 3313 KiB

Open AccessArticle

Preliminary Experimental Quantification of Helium Leakages from Flanged Connections at HCPB TBS Operative Conditions

by Alessandro Venturini, Francesca Papa and Marco Utili

Energies 2023, 16(14), 5519; https://doi.org/10.3390/en16145519 - 21 Jul 2023

Cited by 1 | Viewed by 717

Abstract

The HCPB TBS (Helium-Cooled Pebble Bed Test Blanket System) is one of the two European TBSs that will be installed and tested in the ITER reactor. The use of flanged connections in the Helium Coolant System and the Tritium Extraction System of the [...] Read more.

The HCPB TBS (Helium-Cooled Pebble Bed Test Blanket System) is one of the two European TBSs that will be installed and tested in the ITER reactor. The use of flanged connections in the Helium Coolant System and the Tritium Extraction System of the HCPB TBS would make the remote maintenance operations easier and faster. Therefore, investigating the helium leakage from flanges becomes a fundamental step toward the control of the tritium activity in the Port Cell, as the helium flow will contain a variable but not negligible amount of tritium. The first set of experiments on helium leakages from flanged connections is described in this paper. The experiments were performed in a HeFUS3 facility, an eight-shaped helium loop designed to work at HCPB-TBS-relevant conditions. The facility can provide a helium mass flow rate in the range of 0.27–1.4 kg/s and can reach a pressure as high as 80 bar and a temperature up to 530 °C. Two types of gaskets were tested in this campaign: a spiral-wound gasket and an oval ring joint. The gasket/flange assemblies are described in detail in this paper, together with the test section that hosts them and the performed commissioning tests. The tests were carried out at 500 °C and 80 bar. In these conditions, the leak rate from the flange with the oval ring joint resulted in being, on average, 1.42·10⁻⁶ mbar∙L/s, while the leak rate from the flange with the spiral-wound gasket resulted in being, on average, 3.73·10⁻³ mbar∙L/s. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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12 pages, 10769 KiB

Open AccessArticle

Manufacturing of PAV-ONE, a Permeator against Vacuum Mock-Up with Niobium Membrane

by Francesca Papa, Alessandro Venturini, Gianfranco Caruso, Serena Bassini, Chiara Ciantelli, Angela Fiore, Vincenzo Cuzzola, Antonio Denti and Marco Utili

Energies 2023, 16(14), 5471; https://doi.org/10.3390/en16145471 - 19 Jul 2023

Cited by 2 | Viewed by 751

Abstract

The Permeator Against Vacuum (PAV) is one of the proposed technologies for the Tritium Extraction System of the WCLL BB (Water-Cooled Lithium-Lead Breeding Blanket) of the EU DEMO reactor. In this paper, the manufacturing of the first PAV mock-up with a niobium membrane [...] Read more.

The Permeator Against Vacuum (PAV) is one of the proposed technologies for the Tritium Extraction System of the WCLL BB (Water-Cooled Lithium-Lead Breeding Blanket) of the EU DEMO reactor. In this paper, the manufacturing of the first PAV mock-up with a niobium membrane with a cylindrical configuration is presented. This work aimed to demonstrate the possibility of manufacturing a relevant-size PAV to be later tested in the TRIEX-II facility. The adopted prototypical solutions are described in detail, starting with the methodology developed to join the Nb tubes with a 10CrMo9-10 (A182 F22) plate. Dedicated manufacturing and welding procedures, based on vacuum brazing with a nickel-based brazing alloy, were developed to solve the problem. This new kind of brazing was first analyzed to check the morphology of the joint and then tested to check its capability to withstand the TRIEX-II operative conditions. In parallel, the compatibility with a lithium-lead environment was analyzed by exposing samples of niobium and 10CrMo9-10 (A335 P22) to a flow of the eutectic alloy at 500 °C up to 4000 h. Finally, the PAV mock-up was installed in the TRIEX-II facility. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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27 pages, 12074 KiB

Open AccessArticle

Design and Integration of the WCLL Tritium Extraction and Removal System into the European DEMO Tokamak Reactor

by Marco Utili, Ciro Alberghi, Roberto Bonifetto, Luigi Candido, Aldo Collaku, Belit Garcinuño, Michal Kordač, Daniele Martelli, Rocco Mozzillo, Francesca Papa, David Rapisarda, Laura Savoldi, Fernando R. Urgorri, Domenico Valerio and Alessandro Venturini

Energies 2023, 16(13), 5231; https://doi.org/10.3390/en16135231 - 07 Jul 2023

Cited by 2 | Viewed by 1239

Abstract

The latest progress in the design of the water-cooled lithium–lead (WCLL) tritium extraction and removal (TER) system for the European DEMO tokamak reactor is presented. The implementation and optimization of the conceptual design of the TER system are performed in order to manage [...] Read more.

The latest progress in the design of the water-cooled lithium–lead (WCLL) tritium extraction and removal (TER) system for the European DEMO tokamak reactor is presented. The implementation and optimization of the conceptual design of the TER system are performed in order to manage the tritium concentration in the LiPb and ancillary systems, to control the LiPb chemistry, to remove accumulated corrosion and activated products (in particular, the helium generated in the BB), to store the LiPb, to empty the BB segments, to shield the equipment due to LiPb activation, and to accommodate possible overpressure of the LiPb. The LiPb volumes in the inboard (IB) and outboard (OB) modules of the BB are separately managed due to the different pressure drops and required mass flow rates in the different plasma operational phases. Therefore, the tritium extraction is managed by 6 LiPb loops: 4 loops for the OB segments and 2 loops for the IB segments. Each one is a closed loop with forced circulation of the liquid metal through the TER and the other ancillary systems. The design presents the new CAD drawings and the integration of the TEU into the tokamak building, designed on the basis of an experimental characterization carried out for the permeator against vacuum (PAV) and gas–liquid contactor (GLC) technologies, the two most promising technologies for tritium extraction from liquid metal. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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12 pages, 3900 KiB

Open AccessArticle

European DEMO Fusion Reactor: Design and Integration of the Breeding Blanket Feeding Pipes

by Rocco Mozzillo, Christian Vorpahl, Christian Bachmann, Francisco A. Hernández and Alessandro Del Nevo

Energies 2023, 16(13), 5058; https://doi.org/10.3390/en16135058 - 29 Jun 2023

Cited by 3 | Viewed by 911

Abstract

This article describes the design and configuration of the DEMO Breeding Blanket (BB) feeding pipes inside the upper port. As large BB segments require periodic replacement via the upper vertical ports, the space inside the upper port needs to be maximized. At the [...] Read more.

This article describes the design and configuration of the DEMO Breeding Blanket (BB) feeding pipes inside the upper port. As large BB segments require periodic replacement via the upper vertical ports, the space inside the upper port needs to be maximized. At the same time, the size of the upper port is constrained by the available space in between the toroidal field coils and the required space to integrate a thermal shield between the vacuum vessel (VV) port and the coils. The BB feeding pipes inside the vertical port need to be removed prior to BB maintenance, as they obstruct the removal kinematics. Since they are connected to the BB segments on the top and far from their vertical support on the bottom, the pipes need to be sufficiently flexible to allow for the thermal expansion of the BB segments and the pipes themselves. The optimization and verification of these BB pipes inside the upper port design are critical aspects in the development of DEMO. This article presents the chosen pipe configuration for both BB concepts considered for DEMO (helium- and water-cooled) and their structural verification for some of the most relevant thermal conditions. A 3D model of the pipes forest, both for the Helium-Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL) concepts, has been developed and integrated inside the DEMO Upper Port (UP), Upper Port Ring Channel, and Upper Port Annex (UPA). A preliminary structural analysis of the pipeline was carried out to check the structural integrity of the pipes, their flexibility against the thermal load, their internal pressure, and the deflection induced by the thermal expansion of the BB segments. The results showed that the secondary stress on the hot leg of the HCPB pipeline was above the limit, suggesting future improvements in its shape to increase the flexibility. Moreover, the WCLL concept did not have a critical point in terms of the secondary stress on the pipeline, since the thicknesses and the diameters of these pipes were smaller than the HCPB ones. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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19 pages, 8660 KiB

Open AccessArticle

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

by Alessandra Vannoni, Pierdomenico Lorusso, Marica Eboli, Fabio Giannetti, Cristiano Ciurluini, Amelia Tincani, Ranieri Marinari, Andrea Tarallo and Alessandro Del Nevo

Energies 2023, 16(9), 3729; https://doi.org/10.3390/en16093729 - 26 Apr 2023

Cited by 8 | Viewed by 1697

Abstract

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, [...] Read more.

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. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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12 pages, 7845 KiB

Open AccessArticle

High Heat Flux Testing of Graded W-Steel Joining Concepts for the First Wall

by Vishnu Ganesh, Daniel Dorow-Gerspach, Martin Bram, Christian Linsmeier, Jiri Matejicek and Monika Vilemova

Energies 2023, 16(9), 3664; https://doi.org/10.3390/en16093664 - 24 Apr 2023

Cited by 2 | Viewed by 1152

Abstract

The realization of the first wall (FW), which is composed of a protective tungsten (W) armor covering the structural steel material, is a critical challenge in the development of future fusion reactors. Due to the different coefficients of thermal expansion (CTE) of W [...] Read more.

The realization of the first wall (FW), which is composed of a protective tungsten (W) armor covering the structural steel material, is a critical challenge in the development of future fusion reactors. Due to the different coefficients of thermal expansion (CTE) of W and steel, the direct joining of them results in cyclic thermal stress at their bonding seam during the operation of the fusion reactor. To address this issue, this study benchmarks two joining concepts. The first concept uses an atmospheric plasma sprayed graded interlayer composed of W/steel composites with a varying content of W and steel to gradually change the CTE. The second concept uses a spark plasma sintered graded interlayer. Furthermore, in order to benchmark these concepts, a directly bonded W-steel reference joint as well as a W-steel joint featuring a vanadium interlayer were also tested. These joints were tested under steady-state high heat flux cyclic loading, starting from a heat flux of 1 MW/m² up to 4.5 MW/m², with stepwise increments of 0.5 MW/m². At each heat flux level, 200 thermal cycles were performed. The joints featuring a sintered graded interlayer survived only until 1.5 MW/m² of loading, while the joint featuring plasma sprayed graded interlayer and V interlayer survived until 3 MW/m². Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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10 pages, 3071 KiB

Open AccessArticle

Tritium Extraction from Lithium–Lead Eutectic Alloy: Experimental Characterization of a Permeator against Vacuum Mock-Up at 450 °C

by Francesca Papa, Alessandro Venturini, Daniele Martelli and Marco Utili

Energies 2023, 16(7), 3022; https://doi.org/10.3390/en16073022 - 26 Mar 2023

Cited by 3 | Viewed by 1400

Abstract

Tritium extraction is one of the key open issues toward the development of the WCLL BB (Water-Cooled Lithium–Lead Breeding Blanket) of EU DEMO reactors, and different technologies have been proposed to address it. Among them, the Permeator Against Vacuum (PAV) has promising features, [...] Read more.

Tritium extraction is one of the key open issues toward the development of the WCLL BB (Water-Cooled Lithium–Lead Breeding Blanket) of EU DEMO reactors, and different technologies have been proposed to address it. Among them, the Permeator Against Vacuum (PAV) has promising features, but it has never been tested in a relevant environment. This work presents the first experimental results ever obtained for a PAV mock-up. The experiments were carried out at ENEA Brasimone R.C. with the TRIEX-II facility on a mock-up characterized by a shell and tube configuration and using niobium as a membrane material. The experimental campaign was carried out with LiPb flowing at about 450 °C and 1.2 kg/s, while the hydrogen partial pressure was varied in the range 170–360 Pa. The characterization of the PAV performance was conducted by measuring the hydrogen partial pressure drop across the mock-up and the hydrogen permeated flux through a leak detector calibrated with an external hydrogen calibration cylinder. Moreover, the permeated flux was confirmed by a pressurization test performed measuring the pressure increase on the vacuum side of the PAV. The results constitute the first verification of the possibility to operate a PAV in flowing LiPb and to quantify its capabilities. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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11 pages, 1975 KiB

Open AccessArticle

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

by Amelia Tincani, Cristiano Ciurluini, Alessandro Del Nevo, Fabio Giannetti, Andrea Tarallo, Claudio Tripodo, Antonio Cammi, Alessandra Vannoni, Marica Eboli, Tommaso Del Moro, Pierdomenico Lorusso and Luciana Barucca

Energies 2023, 16(6), 2601; https://doi.org/10.3390/en16062601 - 09 Mar 2023

Cited by 8 | Viewed by 1658

Abstract

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, [...] Read more.

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. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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20 pages, 8598 KiB

Open AccessFeature PaperArticle

Passive Hydrogen Recombination during a Beyond Design Basis Accident in a Fusion DEMO Plant

by Matteo D’Onorio, Tommaso Glingler, Guido Mazzini, Maria Teresa Porfiri and Gianfranco Caruso

Energies 2023, 16(6), 2569; https://doi.org/10.3390/en16062569 - 08 Mar 2023

Cited by 2 | Viewed by 1086

Abstract

One of the most important environmental and safety concerns in nuclear fusion plants is the confinement of radioactive substances into the reactor buildings during both normal operations and accidental conditions. For this reason, hydrogen build-up and subsequent ignition must be avoided, since the [...] Read more.

One of the most important environmental and safety concerns in nuclear fusion plants is the confinement of radioactive substances into the reactor buildings during both normal operations and accidental conditions. For this reason, hydrogen build-up and subsequent ignition must be avoided, since the pressure and energy generated may threaten the integrity of the confinement structures, causing the dispersion of radioactive and toxic products toward the public environment. Potentially dangerous sources of hydrogen are related to the exothermal oxidation reactions between steam and plasma-facing components or hot dust, which could occur during accidents such as the in-vessel loss of coolant or a wet bypass. The research of technical solutions to avoid the risk of a hydrogen explosion in large fusion power plants is still in progress. In the safety and environment work package of the EUROfusion consortium, activities are ongoing to study solutions to mitigate the hydrogen explosion risk. The main objective is to preclude the occurrence of flammable gas mixtures. One identified solution could deal with the installation of passive autocatalytic recombiners into the atmosphere of the vacuum vessel pressure suppression system tanks. A model to control the PARs recombination capacity as a function of thermal-hydraulic parameters of suppression tanks has been modeled in MELCOR. This paper aims to test the theoretical effectiveness of the PAR intervention during an in-vessel loss of coolant accident without the intervention of the decay heat removal system for the Water-Cooled LithiumLead concept of EU-DEMO. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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13 pages, 4383 KiB

Open AccessArticle

Design of a Test Section for the Experimental Investigation of the WCLL Manifold Hydraulic Features

by Aldo Collaku, Pietro Arena, Alessandro Del Nevo, Ranieri Marinari and Laura Savoldi

Energies 2023, 16(5), 2246; https://doi.org/10.3390/en16052246 - 26 Feb 2023

Cited by 3 | Viewed by 1132

Abstract

A scaled-down test section representative of an Outboard Segment manifold of the Water-Cooled Lithium Lead Breeding Blanket for the European DEMO has been designed for installation and test in a high- mass flow branch of the W-HYDRA facility, under construction at the premises [...] Read more.

A scaled-down test section representative of an Outboard Segment manifold of the Water-Cooled Lithium Lead Breeding Blanket for the European DEMO has been designed for installation and test in a high- mass flow branch of the W-HYDRA facility, under construction at the premises of ENEA Brasimone Research Center. The test section should confirm the flow repartition recently computed in the different breeding units on the full-scale manifold, validating at the same time the computational tools used for the design and analysis. The detailed objectives and requirements of the test section, as well as the scaling rationale and procedure adopted for its design, are presented in the paper. The final design of the test section is discussed. The preliminary analyses of the developed design are also presented and show that it is compliant with the initial objectives. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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27 pages, 21380 KiB

Open AccessArticle

Design and Integration of the EU-DEMO Water-Cooled Lead Lithium Breeding Blanket

by Pietro Arena, Gaetano Bongiovì, Ilenia Catanzaro, Cristiano Ciurluini, Aldo Collaku, Alessandro Del Nevo, Pietro Alessandro Di Maio, Matteo D’Onorio, Fabio Giannetti, Vito Imbriani, Pietro Maccari, Lorenzo Melchiorri, Fabio Moro, Rocco Mozzillo, Simone Noce, Laura Savoldi, Simone Siriano, Alessandro Tassone and Marco Utili

Energies 2023, 16(4), 2069; https://doi.org/10.3390/en16042069 - 20 Feb 2023

Cited by 8 | Viewed by 1622

Abstract

The water-cooled lead lithium breeding blanket (WCLL BB) is one of two BB candidate concepts to be chosen as the driver blanket of the EU-DEMO fusion reactor. Research activities carried out in the past decade, under the umbrella of the EUROfusion consortium, have [...] Read more.

The water-cooled lead lithium breeding blanket (WCLL BB) is one of two BB candidate concepts to be chosen as the driver blanket of the EU-DEMO fusion reactor. Research activities carried out in the past decade, under the umbrella of the EUROfusion consortium, have allowed a quite advanced reactor architecture to be achieved. Moreover, significant efforts have been made in order to develop the WCLL BB pre-conceptual design following a holistic approach, identifying interfaces between components and systems while respecting a system engineering approach. This paper reports a description of the current WCLL BB architecture, focusing on the latest modifications in the BB reference layout aimed at evolving the design from its pre-conceptual version into a robust conceptual layout. In particular, the main rationale behind design choices and the BB’s overall performances are highlighted. The present paper also gives an overview of the integration between the BB and the different in-vessel systems interacting with it. In particular, interfaces with the tritium extraction and removal (TER) system and the primary heat transfer system (PHTS) are described. Attention is also paid to auxiliary systems devoted to heat the plasma, such as electron cyclotron heating (ECH). Indeed, the integration of this system in the BB will strongly impact the segment design since it envisages the introduction of significant cut-outs in the BB layout. A preliminary CAD model of the central outboard blanket (COB) segment housing the ECH cut-out has been set up and is reported in this paper. The chosen modeling strategy, adopted loads and boundary conditions, as well as obtained results, are reported in the paper and critically discussed. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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12 pages, 3848 KiB

Open AccessArticle

Design of a Prototypical Mock-Up for the Experimental Investigation of WCLL First-Wall Performances

by Pietro Maccari, Pietro Arena, Ranieri Marinari, Amelia Tincani and Alessandro Del Nevo

Energies 2023, 16(4), 1685; https://doi.org/10.3390/en16041685 - 08 Feb 2023

Cited by 1 | Viewed by 979

Abstract

A large research effort is currently ongoing within the framework of the EUROfusion consortium for the study and design of a water-cooled lithium–lead (WCLL) breeding blanket (BB). This concept will be tested in ITER through the installation of a test blanket module (TBM) [...] Read more.

A large research effort is currently ongoing within the framework of the EUROfusion consortium for the study and design of a water-cooled lithium–lead (WCLL) breeding blanket (BB). This concept will be tested in ITER through the installation of a test blanket module (TBM) and it is one of the two candidates adopted as driver BBs in DEMO. In this framework, at the ENEA research centre of Brasimone, the realization of the experimental platform, W-HYDRA, is envisaged. The platform is dedicated to the support of the development of WCLL BB and ITER TBM and the investigation of the DEMO balance of plants. One of the most important experimental infrastructures is the water-loop facility, the aim of which is to provide water at a high pressure and temperature (PWR conditions), with a sufficient mass-flow rate and power for the experimental testing of BB and TBM components. The facility will be equipped with a vacuum chamber and an electron beam gun for the reproduction of high surface heat flux on plasma-facing components. In the present work, the design of a prototypical mock-up (MU) of the WCLL BB first wall is described. The MU is used to investigate the thermal, hydraulic and structural behavior of the current first-wall design under relevant heat loads at the expected operational conditions. The delineation of the main experimental test’s features and the instrumentation needed is assessed in the paper. A preliminary CFD calculation on the prototypical MU and the computational results are also presented. Full article

(This article belongs to the Special Issue Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022)

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Water Coolant Lithium Lead Breeding Blanket and Test Blanket Module: Design and R&D Activities in 2022

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