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Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications

A special issue of Energies (ISSN 1996-1073). This special issue belongs to the section "J: Thermal Management".

Deadline for manuscript submissions: closed (31 October 2021) | Viewed by 34699

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Special Issue Editors

Italian National Agency for New Technologies, Energy and Sustainable Economic Development, ENEA FSN-ING-SIS, CR Brasimone, 40032 Camugnano, BO, Italy
Interests: nuclear energy; nuclear engineering; fusion technologies; water-cooled lithium–lead breeding blanket technology; design and conduction of experimental facilities
Special Issues, Collections and Topics in MDPI journals
Italian National Agency for New Technologies, Energy and Sustainable Economic Development, ENEA FSN-ING-SIS, CR Brasimone, 40032 Camugnano, BO, Italy
Interests: nuclear energy; nuclear engineering; thermal-hydraulics, fusion technologies; water cooled lithium lead breeding blanket technology; system code development and validation; design and conduction of experiemental facilities

Special Issue Information

Dear Colleagues,

The perspective of having an almost inexhaustible source of energy has driven worldwide research initiatives on nuclear fusion technology, which have the main representation with the construction of the international nuclear fusion research and engineering megaproject (ITER) in Europe. In this framework, thermal hydraulics is a key discipline which is essential for design and safety demands. Thermal hydraulics is employed in the design phase of the systems and components to demonstrate the performances, to ensure the reliability, and to guarantee efficient and economical operation. It is in charge of investigating the transients of the engineering systems: this includes safety analysis aimed at demonstrating that the operation is safe and consistent with the regulatory authority requirements.

In nuclear fusion technology, thermal hydraulics is required for the design and analysis of cooling and ancillary systems, such as the blanket (i.e. breeding, test or shielding), the divertor, the cryogenic, and the balance of plant systems, as well as the tritium carrier, extraction and recovery systems. Thus, the analyses involve different fluids (water, non-condensable gases, liquid metals) in a broad spectrum of operative conditions (from cryogenic to high temperatures, from vacuum to high pressures), which make their compatibility with the materials challenging.

Although the knowledge of thermal hydraulics benefits from the tremendous efforts conducted for the nuclear fission technology development, it still has remarkable gaps when fusion technology needs are concerned.

Numerical tools, such as well-established system codes, suffer because fluids, parameter ranges, and a field of applications are outside their development and validation boundaries. More sophisticated and complex CFD codes are challenged by the geometrical dimension and complexity of the domains, and by the multi-physics requirements of the analyses.

A considerable amount of resources is devoted at the international level for constructing experimental infrastructures, and for establishing and conducting experimental programs, in full scale and scaled-down facilities, which are aimed at demonstrating the technical feasibility of system and component designs, as well as at generating reference databases to support code development and validation.

In view of the above, it has been proposed to collect the technical status and challenges, and to document the recent scientific advancements of “Thermal-hydraulics in Nuclear Fusion” in a Special Issue. Scientific and technical contributions include, but are not limited to:

  • Thermal hydraulic analyses of systems and components, including magneto-hydrodynamics;
  • Safety investigations of systems and components;
  • Numerical models development;
  • Code development and application;
  • Codes coupling methodology;
  • Code assessment and validation, including benchmarks;
  • Experimental infrastructures design and operation;
  • Experimental campaigns and investigations;
  • Scaling issue in experiments.

Papers are invited from experts working in areas mentioned above for publication in the Special Issue.

Dr. Alessandro Del Nevo
Dr. Marica Eboli
Guest Editors

Manuscript Submission Information

Manuscripts should be submitted online at www.mdpi.com by registering and logging in to this website. Once you are registered, click here to go to the submission form. Manuscripts can be submitted until the deadline. All submissions that pass pre-check are peer-reviewed. Accepted papers will be published continuously in the journal (as soon as accepted) and will be listed together on the special issue website. Research articles, review articles as well as short communications are invited. For planned papers, a title and short abstract (about 100 words) can be sent to the Editorial Office for announcement on this website.

Submitted manuscripts should not have been published previously, nor be under consideration for publication elsewhere (except conference proceedings papers). All manuscripts are thoroughly refereed through a single-blind peer-review process. A guide for authors and other relevant information for submission of manuscripts is available on the Instructions for Authors page. Energies is an international peer-reviewed open access semimonthly journal published by MDPI.

Please visit the Instructions for Authors page before submitting a manuscript. The Article Processing Charge (APC) for publication in this open access journal is 2600 CHF (Swiss Francs). Submitted papers should be well formatted and use good English. Authors may use MDPI's English editing service prior to publication or during author revisions.

Keywords

  • Nuclear Fusion
  • Breeding blanket system
  • Divertor system
  • Vacuum vessel system
  • Balance of plant
  • Cryogenic system
  • Thermal-hydraulics
  • Two phase flow
  • Heat transfer
  • High flux components
  • Thermal hydraulics of cryogenic fluids
  • Magnetohydrodynamics
  • Experiments
  • Numerical simulations
  • Accident analysis
  • Verification and validation
  • Numerical models development

Published Papers (20 papers)

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Research

30 pages, 13000 KiB  
Article
Experimental and Numerical Results of LIFUS5/Mod3 Series E Test on In-Box LOCA Transient for WCLL-BB
Energies 2021, 14(24), 8527; https://doi.org/10.3390/en14248527 - 17 Dec 2021
Cited by 9 | Viewed by 1832
Abstract
The in-box LOCA (Loss of Coolant Accident) represents a major safety concern to be addressed in the design of the WCLL-BB (water-cooled lead-lithium breeding blanket). Research activities are ongoing to master the phenomena and processes that occur during the postulated accident, to enhance [...] Read more.
The in-box LOCA (Loss of Coolant Accident) represents a major safety concern to be addressed in the design of the WCLL-BB (water-cooled lead-lithium breeding blanket). Research activities are ongoing to master the phenomena and processes that occur during the postulated accident, to enhance the predictive capability and reliability of numerical tools, and to validate computer models, codes, and procedures for their applications. Following these objectives, ENEA designed and built the new separate effects test facility LIFUS5/Mod3. Two experimental campaigns (Series D and Series E) were executed by injecting water at high pressure into a pool of PbLi in WCLL-BB-relevant parameter ranges. The obtained experimental data were used to check the capabilities of the RELAP5 system code to reproduce the pressure transient of a water system, to validate the chemical model of PbLi/water reactions implemented in the modified version of SIMMER codes for fusion application, to investigate the dynamic effects of energy release on the structures, and to provide relevant feedback for the follow-up experimental campaigns. This work presents the experimental data and the numerical simulations of Test E4.1. The results of the test are presented and critically discussed. The code simulations highlight that SIMMER code is able to reproduce the phenomena connected to PbLi/water interaction, and the relevant test parameters are in agreement with the acquired experimental signals. Moreover, the results obtained by the first approach to SIMMER-RELAP5 code-coupling demonstrate its capability of and strength for predicting the transient scenario in complex geometries, considering multiple physical phenomena and minimizing the computational cost. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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17 pages, 12879 KiB  
Article
Hydraulic Characterization of the Full Scale Mock-Up of the DEMO Divertor Outer Vertical Target
Energies 2021, 14(23), 8086; https://doi.org/10.3390/en14238086 - 02 Dec 2021
Cited by 4 | Viewed by 1473
Abstract
In the frame of the pre-conceptual design activities of the DEMO work package DIV-1 “Divertor Cassette Design and Integration” of the EUROfusion program, a mock-up of the divertor outer vertical target (OVT) was built, mainly in order to: (i) demonstrate the technical feasibility [...] Read more.
In the frame of the pre-conceptual design activities of the DEMO work package DIV-1 “Divertor Cassette Design and Integration” of the EUROfusion program, a mock-up of the divertor outer vertical target (OVT) was built, mainly in order to: (i) demonstrate the technical feasibility of manufacturing procedures; (ii) verify the hydraulic design and its capability to ensure a uniform and proper cooling for the plasma facing units (PFUs) with an acceptable pressure drop; and (iii) experimentally validate the computational fluid-dynamic (CFD) model developed by the University of Palermo. In this context, a research campaign was jointly carried out by the University of Palermo and ENEA to experimentally and theoretically assess the hydraulic performances of the OVT mock-up, paying particular attention to the coolant distribution among the PFUs and the total pressure drop across the inlet and outlet sections of the mock-up. The paper presents the results of the steady-state hydraulic experimental test campaign performed at ENEA Brasimone Research Center as well as the relevant numerical analyses performed at the Department of Engineering at the University of Palermo. The test facility, the experimental apparatus, the test matrix and the experimental results, as well as the theoretical model, its assumptions, and the analyses outcomes are herewith reported and critically discussed. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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15 pages, 3005 KiB  
Article
A Validation Roadmap of Multi-Physics Simulators of the Resonator of MW-Class CW Gyrotrons for Fusion Applications
Energies 2021, 14(23), 8027; https://doi.org/10.3390/en14238027 - 01 Dec 2021
Cited by 7 | Viewed by 1586
Abstract
For a few years the multi-physics modelling of the resonance cavity (resonator) of MW-class continuous-wave gyrotrons, to be employed for electron cyclotron heating and current drive in magnetic confinement fusion machines, has gained increasing interest. The rising target power of the gyrotrons, which [...] Read more.
For a few years the multi-physics modelling of the resonance cavity (resonator) of MW-class continuous-wave gyrotrons, to be employed for electron cyclotron heating and current drive in magnetic confinement fusion machines, has gained increasing interest. The rising target power of the gyrotrons, which drives progressively higher Ohmic losses to be removed from the resonator, together with the need for limiting the resonator deformation as much as possible, has put more emphasis on the thermal-hydraulic and thermo-mechanic modeling of the cavity. To cope with that, a multi-physics simulator has been developed in recent years in a shared effort between several European institutions (the Karlsruher Institut für Technologie and Politecnico di Torino, supported by Fusion for Energy). In this paper the current status of the tool calibration and validation is addressed, aiming at highlighting where any direct or indirect comparisons with experimental data are missing and suggesting a possible roadmap to fill that gap, taking advantage of forthcoming tests in Europe. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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13 pages, 3505 KiB  
Article
Overview of Thermal Hydraulic Optimization and Verification for the EU-DEMO HCPB BOP ICD Variant
Energies 2021, 14(23), 7894; https://doi.org/10.3390/en14237894 - 25 Nov 2021
Cited by 5 | Viewed by 1710
Abstract
When progressing from the International Thermonuclear Experimental Reactor (ITER) to the Demonstration Fusion Reactor (DEMO), a system for transferring plasma heat exhaust to a power conversion system is necessary for the so-called Balance of Plant (BOP). During the preconceptual phase of the EU-DEMO [...] Read more.
When progressing from the International Thermonuclear Experimental Reactor (ITER) to the Demonstration Fusion Reactor (DEMO), a system for transferring plasma heat exhaust to a power conversion system is necessary for the so-called Balance of Plant (BOP). During the preconceptual phase of the EU-DEMO project, different BOP concepts were investigated in order to identify the main requirements and feasible architectures to achieve that goal in the most efficient way. This paper comprises the investigations performed during the DEMO preconceptual design phase (p-CDP) and compares the different variants. The main aspect was focused on the helium-cooled pebble bed (HCPB) breeding blanket (BB) concept. After all assessments were performed, the indirect coupled design (ICD) was chosen as the reference configuration for the DEMO HCPB BOP for further development and optimization. The ICD provides decoupling using a molten salt storage loop, which accumulates thermal power during plasma pulses that are released during dwell periods. The work is supported by simulations using design codes EBSILON and MATLAB/SIMULINK, providing the basis for the next design phase. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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22 pages, 39960 KiB  
Article
Experimental Investigation of EU-DEMO Breeding Blanket First Wall Mock-Ups in Support of the Manufacturing and Material Development Programmes
Energies 2021, 14(22), 7580; https://doi.org/10.3390/en14227580 - 12 Nov 2021
Cited by 2 | Viewed by 1269
Abstract
This paper presents the testing campaign of the two First Wall mock-ups in the HELOKA facility, one mock-up having a 3 mm thick Oxide Dispersion Strengthened (ODS) steel layer on its surface and the other featuring a tungsten functionally graded cover. Special consideration [...] Read more.
This paper presents the testing campaign of the two First Wall mock-ups in the HELOKA facility, one mock-up having a 3 mm thick Oxide Dispersion Strengthened (ODS) steel layer on its surface and the other featuring a tungsten functionally graded cover. Special consideration is given to the diagnostics used for these tests, in particular, the measurement of the surface temperature of the tungsten functionally graded layer with an infrared camera. Additionally, the paper looks into the uncertainty associated with the calorimetric evaluation of the applied heating power for these experiments. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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14 pages, 6203 KiB  
Article
A Water Loop Design for the CRAFT Project towards the Testing of CFETR Water-Cooled Blanket and Divertor
Energies 2021, 14(21), 7354; https://doi.org/10.3390/en14217354 - 04 Nov 2021
Viewed by 1467
Abstract
As one of the tasks of the Comprehensive Research Facility for Fusion Technology (CRAFT), a High Heat Flux (HHF) testing device will be built to test the blanket and divertor of Chinese Fusion Engineering Testing Reactor (CFETR). The water loop is a key [...] Read more.
As one of the tasks of the Comprehensive Research Facility for Fusion Technology (CRAFT), a High Heat Flux (HHF) testing device will be built to test the blanket and divertor of Chinese Fusion Engineering Testing Reactor (CFETR). The water loop is a key system of the HHF testing device. The main objective of the water loop is to provide deionized water at specific temperature, pressure, and flow rate for different testing experiments of the water-cooled blanket and water-cooled divertor components. The design of the water loop has been through three major steps. Firstly, the water cooled blanket and divertor were designed and analyzed, in detail, for CFETR. Secondly, thermal hydraulic features of the prototypes were abstracted from the analyses results. Then, the experiment plan was made so that the preliminary design of the water loop was carried out. The third step was the engineering design, which was conducted through cooperation with an industrial enterprise with certifications. At present, the water loop is ready for fabrication and construction. The water loop will be completed, for commissioning operation, by August 2022, as scheduled. After that, the experiments will be carried out step by step and provide solid technical base to CFETR. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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16 pages, 12237 KiB  
Article
Model Development and Transient Analysis of the HCPB BB BOP DEMO Configuration Using the Apros System Code
Energies 2021, 14(21), 7214; https://doi.org/10.3390/en14217214 - 02 Nov 2021
Cited by 1 | Viewed by 1337
Abstract
Extensive modeling and analytical work has been carried out considering the Helium-Cooled Pebble Bed Breeding Blanket (HCPB BB) Balance Of Plant (BOP) configuration of the Demonstration Power Plant (DEMO) using the Apros system code, developed by VTT Technical Research Centre of Finland Ltd. [...] Read more.
Extensive modeling and analytical work has been carried out considering the Helium-Cooled Pebble Bed Breeding Blanket (HCPB BB) Balance Of Plant (BOP) configuration of the Demonstration Power Plant (DEMO) using the Apros system code, developed by VTT Technical Research Centre of Finland Ltd. and Fortum. The integral plant model of the HCPB BB plant has been improved with respect to the blanket and steam generator models. Based on HCPB-BL2017 v1 data, reported in 2019, the blanket has been remodeled by separate Apros process components, dedicated to average inboard and outboard segments, where the power deposition scheme of the breeding units took into account the output of high-fidelity neutronic analyses. A new helical coil steam generator model has been developed for primary–secondary system coupling using CAD data provided by EUROfusion partner University of Palermo. Transient analyses have been performed with Apros on the plant configuration that utilizes a molten salt technology-based small Energy Storage System (ESS). Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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19 pages, 6029 KiB  
Article
Numerical Simulation of an Out-Vessel Loss of Coolant from the Breeder Primary Loop Due to Large Rupture of Tubes in a Primary Heat Exchanger in the DEMO WCLL Concept
Energies 2021, 14(21), 6916; https://doi.org/10.3390/en14216916 - 21 Oct 2021
Viewed by 1328
Abstract
This work presents a thermohydraulic analysis of a postulated accident involving the rupture of the breeder primary cooling loop inside a heat exchanger (once through steam generator). After the detection of the loss of pressure inside the primary loop, a plasma shutdown is [...] Read more.
This work presents a thermohydraulic analysis of a postulated accident involving the rupture of the breeder primary cooling loop inside a heat exchanger (once through steam generator). After the detection of the loss of pressure inside the primary loop, a plasma shutdown is actuated with a consequent plasma disruption, isolation of the secondary loop, and shutoff of the pumps in the primary; no other safety counteractions are postulated. The objective of the work is to analyze the pressurization of the primary and secondary sides to show that the accidental overpressure in the two sides of the steam generators is safely accommodated. Furthermore, the effect of the plasma disruption on the FW, in terms of temperatures, should be analyzed. Lastly, the time transients of the pressures and temperatures in the HX and BB for a time span of up to 36 h should be obtained to assess the effect of the decay heat over a long period. A full nodalization of the OTSG was realized together with a simplified nodalization of the whole PHTS BB loop. The code utilized was MELCOR for fusion version 1.8.6. The accident was simulated by activating a flow path which directly connected one section of the primary with the parallel section of the secondary side. It is shown here that the pressures and the temperatures inside the whole PHTS system remain below the safety thresholds for the whole transient. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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37 pages, 4830 KiB  
Article
MHD R&D Activities for Liquid Metal Blankets
Energies 2021, 14(20), 6640; https://doi.org/10.3390/en14206640 - 14 Oct 2021
Cited by 13 | Viewed by 2201
Abstract
According to the most recently revised European design strategy for DEMO breeding blankets, mature concepts have been identified that require a reduced technological extrapolation towards DEMO and will be tested in ITER. In order to optimize and finalize the design of test blanket [...] Read more.
According to the most recently revised European design strategy for DEMO breeding blankets, mature concepts have been identified that require a reduced technological extrapolation towards DEMO and will be tested in ITER. In order to optimize and finalize the design of test blanket modules, a number of issues have to be better understood that are related to the magnetohydrodynamic (MHD) interactions of the liquid breeder with the strong magnetic field that confines the fusion plasma. The aim of the present paper is to describe the state of the art of the study of MHD effects coupled with other physical phenomena, such as tritium transport, corrosion and heat transfer. Both numerical and experimental approaches are discussed, as well as future requirements to achieve a reliable prediction of these processes in liquid metal blankets. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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15 pages, 4523 KiB  
Article
Loss of Liquid Lithium Coolant in an Accident in a DONES Test Cell Facility
Energies 2021, 14(20), 6569; https://doi.org/10.3390/en14206569 - 12 Oct 2021
Cited by 3 | Viewed by 1454
Abstract
A Demo-Oriented early NEutron Source (DONES) facility for material irradiation with nuclear is currently being designed. DONES aims to produce neutrons with fusion-relevant spectrum and fluence by means of D–Li stripping reactions occurring between a deuteron beam impacting a stable liquid lithium flowing [...] Read more.
A Demo-Oriented early NEutron Source (DONES) facility for material irradiation with nuclear is currently being designed. DONES aims to produce neutrons with fusion-relevant spectrum and fluence by means of D–Li stripping reactions occurring between a deuteron beam impacting a stable liquid lithium flowing film implementing the target. Given the hazard constituted by the liquid lithium inventory and the potential risk of reactions with water, air, and concrete eventually resulting in fire events, the Target Test Cell (TTC) is filled with helium and the reinforced concrete walls forming the bio-shield are covered with steel liners. A loss of Li in TTC, due to a large break in the Quench Tank, is postulated, and consequences are deterministically studied. With the TTC liner being water-cooled, the impact of the liner temperature rise following a leakage event is evaluated. Two separate MELCOR code models have been defined for the liquid lithium loop and water-cooled loop and are numerically coupled. The amount of leaked inventory dependent on the implemented safety logic and impact on TTC containment is evaluated. The water pressurization pattern within the liner cooling loop is studied to highlight possible risks of lithium–water/concrete reactions. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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10 pages, 3735 KiB  
Article
Wake Shape and Height Profile Measurements in a Concave Open Channel Flow regarding the Target in DONES
Energies 2021, 14(20), 6506; https://doi.org/10.3390/en14206506 - 11 Oct 2021
Cited by 1 | Viewed by 1179
Abstract
Wakes appearing downstream of disturbances on the surface of a water flow in a concave open channel were examined experimentally. The investigated channel geometry was similar to the liquid lithium target in DONES (Demonstration fusion power plant Oriented NEutron Source). The objective of [...] Read more.
Wakes appearing downstream of disturbances on the surface of a water flow in a concave open channel were examined experimentally. The investigated channel geometry was similar to the liquid lithium target in DONES (Demonstration fusion power plant Oriented NEutron Source). The objective of the measurements was to analyze the effect of a disturbance on the downstream layer thickness. For measuring the height profiles in the channel, an optical measurement system based on laser triangulation was developed. It was shown that the wake of the undisturbed flow emerged from the nozzle corner, which was in accordance with analytical solutions. For sufficiently large disturbances at the nozzle edge, the height profiles located downstream showed symmetrical minima and maxima on both sides of the disturbance. The wake depth strongly depended on the diameter and penetration depth of the disturbance, as well as the circumferential position in the channel, which yields to a critical wake depth of one millimeter for the lithium target in DONES. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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17 pages, 15485 KiB  
Article
Thermal Hydraulic Analysis on the Water Lead Lithium Cooled Blanket for CFETR
Energies 2021, 14(19), 6350; https://doi.org/10.3390/en14196350 - 05 Oct 2021
Cited by 1 | Viewed by 1341
Abstract
A new type of Water Lead Lithium Cooled (WLLC) blanket that adopts the modular design scheme, water cooling the structure components, liquid PbLi as breeder and coolant, and SiC as the thermal insulator between PbLi and structures is under development as a candidate [...] Read more.
A new type of Water Lead Lithium Cooled (WLLC) blanket that adopts the modular design scheme, water cooling the structure components, liquid PbLi as breeder and coolant, and SiC as the thermal insulator between PbLi and structures is under development as a candidate blanket concept for the Chinese Fusion Engineering Test Reactor (CFETR). Based on a poloidal-radial slice model, thermal hydraulic analysis is performed for this blanket to validate the feasibility of design goals. Results show that the present design can achieve the outlet temperature in the range of 600–700 °C, with all the material temperatures safely below the upper limits. A series of sensitivity analyses are also carried out. It indicates that the thermal conductivity (TC) of SiC would have a significant influence on the temperature field, streamlines and pressure drop; that is, lower TC of SiC can maintain the temperature of PbLi at a high level, and induce an increased number of vortices in the liquid PbLi flow as well as a larger pressure drop. On this basis, the joint effects of the TC of SiC and inlet velocity on the performance of blanket thermal hydraulics are analyzed, then the so-called “attainable region” is proposed. Finally, optimization design studies are carried out by decreasing the width of the front channel. Comparison results show that the present design is the most reasonable. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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16 pages, 2976 KiB  
Article
Magneto-Convective Analyses of the PbLi Flow for the EU-WCLL Fusion Breeding Blanket
Energies 2021, 14(19), 6192; https://doi.org/10.3390/en14196192 - 28 Sep 2021
Cited by 6 | Viewed by 1493
Abstract
The Water Cooled Lithium Lead (WCLL) breeding blanket is one of the driver blanket concepts under development for the European Demonstration Reactor (DEMO). The majority of the blanket volume is occupied by flowing PbLi at eutectic composition. This liquid metal flow is subdued [...] Read more.
The Water Cooled Lithium Lead (WCLL) breeding blanket is one of the driver blanket concepts under development for the European Demonstration Reactor (DEMO). The majority of the blanket volume is occupied by flowing PbLi at eutectic composition. This liquid metal flow is subdued to high fluxes of particles coming from the plasma which are translated into a high non-homogeneous heat volumetric source inside the fluid. The heat is removed from the PbLi thanks to several water tubes immersed in the metal. The dynamics of the PbLi is heavily affected by the heat source and by the position of the tubes. Moreover, the conducting fluid is electrically coupled with the intense magnetic field used for the plasma confinement. As a result, the PbLi flow is strongly affected by the Magnetohydrodynamics (MHD) forces. In the WCLL, the MHD and convective interactions are expected to be comparable. Therefore, the PbLi dynamics and consequently the heat transfer between the liquid metal and the water coolant will be ruled by the magneto-convective phenomenon. This work presents 3D computational analyses of the PbLi flow in the frontal region of the WCLL design. The simulations include the combined effect of MHD forces caused by the magnetic field and the buoyancy interaction created by the temperature distribution. The latter is determined by the PbLi dynamics, the volumetric heat source and the position of the water tubes. Simulations have allowed computing the heat transfer between the PbLi and the water tubes. Nusselt and Grashof numbers have been obtained in the different regions of the system. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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17 pages, 12818 KiB  
Article
Model Development and Transient Analysis of the WCLL BB BOP DEMO Configuration Using the Apros System Code
Energies 2021, 14(18), 5593; https://doi.org/10.3390/en14185593 - 07 Sep 2021
Cited by 4 | Viewed by 1471
Abstract
Extensive modelling and analytical work has been carried out considering the water-cooled lithium–lead breeding blanket (WCLL BB) balance of plant (BOP) configuration of the demonstration power plant (DEMO) using the Apros system code, developed by VTT Technical Research Centre of Finland Ltd. and [...] Read more.
Extensive modelling and analytical work has been carried out considering the water-cooled lithium–lead breeding blanket (WCLL BB) balance of plant (BOP) configuration of the demonstration power plant (DEMO) using the Apros system code, developed by VTT Technical Research Centre of Finland Ltd. and Fortum. Contributing to the BOP work package of the EUROfusion Consortium, the integral plant model for dynamic analyses of the WCLL BB configuration has been updated with special attention to primary system components. Following trends of relevant neutronics modelling, a new BB model has been implemented in 2020 with the aim to obtain higher resolution output data and a more realistic thermalhydraulic feedback from the primary system. Once-through steam generator user components have been built based on CAD models conceived by BOP partners. Transient analyses have been performed providing a better picture regarding the behaviour of main components, e.g., the BB and the OTSGs, whilst highlighting possible ways to optimise the control scheme of the plant. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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29 pages, 2311 KiB  
Article
Development of a RELAP5/MOD3.3 Module for MHD Pressure Drop Analysis in Liquid Metals Loops: Verification and Validation
Energies 2021, 14(17), 5538; https://doi.org/10.3390/en14175538 - 04 Sep 2021
Cited by 12 | Viewed by 1888
Abstract
Magnetohydrodynamic (MHD) phenomena, due to the interaction between a magnetic field and a moving electro-conductive fluid, are crucial for the design of magnetic-confinement fusion reactors and, specifically, for the design of the breeding blanket concepts that adopt liquid metals (LMs) as working fluids. [...] Read more.
Magnetohydrodynamic (MHD) phenomena, due to the interaction between a magnetic field and a moving electro-conductive fluid, are crucial for the design of magnetic-confinement fusion reactors and, specifically, for the design of the breeding blanket concepts that adopt liquid metals (LMs) as working fluids. Computational tools are employed to lead fusion-relevant physical analysis, but a dedicated MHD code able to simulate all the phenomena involved in a blanket is still not available and there is a dearth of systems code featuring MHD modelling capabilities. In this paper, models to predict both 2D and 3D MHD pressure drop, derived by experimental and numerical works, have been implemented in the thermal-hydraulic system code RELAP5/MOD3.3 (RELAP5). The verification and validation procedure of the MHD module involves the comparison of the results obtained by the code with those of direct numerical simulation tools and data obtained by experimental works. As relevant examples, RELAP5 is used to recreate the results obtained by the analysis of two test blanket modules: Lithium Lead Ceramic Breeder and Helium-Cooled Lithium Lead. The novel MHD subroutines are proven reliable in the prediction of the pressure drop for both simple and complex geometries related to LM circuits at high magnetic field intensity (error range ±10%). Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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16 pages, 3551 KiB  
Article
Verification and Validation of COMSOL Magnetohydrodynamic Models for Liquid Metal Breeding Blankets Technologies
Energies 2021, 14(17), 5413; https://doi.org/10.3390/en14175413 - 31 Aug 2021
Cited by 4 | Viewed by 2348
Abstract
Liquid metal breeding blankets are extensively studied in nuclear fusion. In the main proposed systems, the Water Cooled Lithium Lead (WCLL) and the Dual Coolant Lithium Lead (DCLL), the liquid metal flows under an intense transverse magnetic field, for which a magnetohydrodynamic (MHD) [...] Read more.
Liquid metal breeding blankets are extensively studied in nuclear fusion. In the main proposed systems, the Water Cooled Lithium Lead (WCLL) and the Dual Coolant Lithium Lead (DCLL), the liquid metal flows under an intense transverse magnetic field, for which a magnetohydrodynamic (MHD) effect is produced. The result is the alteration of all the flow features and the increase in the pressure drops. Although the latter issue can be evaluated with system models, 3D MHD codes are of extreme importance both in the design phase and for safety analyses. To test the reliability of COMSOL Multiphysics for the development of MHD models, a method for verification and validation of magnetohydrodynamic codes is followed. The benchmark problems solved regard steady state, fully developed flows in rectangular ducts, non-isothermal flows, flow in a spatially varying transverse magnetic field and two different unsteady turbulent problems, quasi-two-dimensional MHD turbulent flow and 3D turbulent MHD flow entering a magnetic obstacle. The computed results show good agreement with the reference solutions for all the addressed problems, suggesting that COMSOL can be used as software to study liquid metal MHD problems under the flow regimes typical of fusion power reactors. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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23 pages, 11601 KiB  
Article
Numerical Simulations with RELAP5-3D and RELAP5/mod3.3 of the Second Experimental Campaign on In-Box LOCA Transients for HCLL TBS
Energies 2021, 14(15), 4544; https://doi.org/10.3390/en14154544 - 27 Jul 2021
Viewed by 1856
Abstract
In-box LOCA was identified as one of the worst accidental scenarios for the HCLL TBS (Helium Cooled Lithium-Lead Test Blanket System). Aiming to experimentally analyze the consequences of this transient, ENEA designed and built THALLIUM (Test HAmmer in Lead LIthiUM), a facility that [...] Read more.
In-box LOCA was identified as one of the worst accidental scenarios for the HCLL TBS (Helium Cooled Lithium-Lead Test Blanket System). Aiming to experimentally analyze the consequences of this transient, ENEA designed and built THALLIUM (Test HAmmer in Lead LIthiUM), a facility that reproduces the LiPb loop of the HCLL TBS. Two experimental campaigns were carried out by simulating the rupture of a stiffening plate and the related helium injection in the LiPb loop. The obtained experimental data were used to check the capabilities of RELAP5 system code to reproduce the pressure wave propagation that follows this accident. The first simulations were made with RELAP5-3D using LBE (Lead–Bismuth Eutectic) as a system fluid, as the thermophysical properties of LiPb are tabulated only up to a maximum value of 40 bar in this version of the code. Then, LiPb properties were implemented in RELAP5/mod3.3, after selecting the proper correlations from a literature review. This work summarizes the numerical simulations of the second experimental campaign, which was simulated with both versions of the code. The simulations highlight that the code is able to accurately reproduce the experimental results and that RELAP5-3D is slightly more precise than RELAP5/mod3.3 in predicting the pressure trends. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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15 pages, 6618 KiB  
Article
CFD Optimization of the Resistivity Meter for the IFMIF-DONES Facility
Energies 2021, 14(9), 2543; https://doi.org/10.3390/en14092543 - 28 Apr 2021
Viewed by 1331
Abstract
A detailed study of lithium-related topics in the IFMIF-DONES facility is currently being promoted and supported within the EUROfusion action, paying attention to different pivotal aspects including lithium flow stability and the monitoring and extraction of impurities. The resistivity meter is a device [...] Read more.
A detailed study of lithium-related topics in the IFMIF-DONES facility is currently being promoted and supported within the EUROfusion action, paying attention to different pivotal aspects including lithium flow stability and the monitoring and extraction of impurities. The resistivity meter is a device able to monitor online non-metallic impurities (mainly nitrogen) in flowing lithium. It relies on the variation of the electric resistivity produced by dissolved anions: the higher the concentration of impurities in lithium, the higher the resistivity measured. The current configuration of the resistivity meter has shown different measuring issues during its operation. All these issues reduce the accuracy of the measurements performed with this instrument and introduce relevant noise affecting the resistance value. This paper proposes different upgrades, supported by CFD simulations, to optimize lithium flow conditions and to reduce measurement problems. Owing to these upgrades, a new design of the resistivity meter has been achieved, which is simpler and easier to manufacture. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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33 pages, 12690 KiB  
Article
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; https://doi.org/10.3390/en14061541 - 11 Mar 2021
Cited by 10 | Viewed by 1787
Abstract
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 [...] Read more.
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. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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22 pages, 11232 KiB  
Article
Effect of Pebble Size Distribution and Wall Effect on Inner Packing Structure and Contact Force Distribution in Tritium Breeder Pebble Bed
Energies 2021, 14(2), 449; https://doi.org/10.3390/en14020449 - 15 Jan 2021
Cited by 8 | Viewed by 2181
Abstract
In the tritium breeding blanket of nuclear fusion reactors, the heat transfer behavior and thermal-mechanical response of the tritium breeder pebble bed are affected by the inner packing structure, which is crucial for the design and optimization of a reliable pebble bed in [...] Read more.
In the tritium breeding blanket of nuclear fusion reactors, the heat transfer behavior and thermal-mechanical response of the tritium breeder pebble bed are affected by the inner packing structure, which is crucial for the design and optimization of a reliable pebble bed in tritium breeding blanket. Thus, the effect of pebble size distribution and fixed wall effect on packing structure and contact force in the poly-disperse pebble bed were investigated by numerical simulation. The results show that pebble size distribution has a significant influence on the inner packing structure of pebble bed. With the increase of the dispersion of pebble size, the average porosity and the average coordination number of the poly-disperse pebble bed gradually decrease. Due to the influence of the fixed wall, the porosity distribution of the pebble bed shows an obvious wall effect. For poly-disperse pebble bed, the influenced region of the wall effect gradually decreases with the increase of the dispersion of pebble size. In addition, the gravity effect and the pebble size distribution have an obvious influence on the contact force distribution inside the poly-disperse pebble bed. The majority of the contact force are weak contact force that is less than the average contact force. Only a few of pebbles have strong contact force that is greater than average contact force. This investigation can help in analyzing the pebble crushing characteristics and the thermal hydraulic analysis in the poly-disperse tritium breeder pebble bed. Full article
(This article belongs to the Special Issue Thermal-Hydraulics in Nuclear Fusion Technology: R&D and Applications)
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