Journal of Nuclear Engineering

Research

37 pages, 17102 KiB

Open AccessArticle

Advancements in Designing the DEMO Driver Blanket System at the EU DEMO Pre-Conceptual Design Phase: Overview, Challenges and Opportunities

by Francisco A. Hernández, Pietro Arena, Lorenzo V. Boccaccini, Ion Cristescu, Alessandro Del Nevo, Pierre Sardain, Gandolfo A. Spagnuolo, Marco Utili, Alessandro Venturini and Guangming Zhou

J. Nucl. Eng. 2023, 4(3), 565-601; https://doi.org/10.3390/jne4030037 - 3 Aug 2023

Cited by 6 | Viewed by 2974

Abstract

The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion [...] Read more.

The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion power plant (FPP) by playing the role of a “Component Test Facility” for the BB. Within this strategy, a so-called driver blanket, with nearly full in-vessel surface coverage, will aim at achieving high-level stakeholder requirements of tritium self-sufficiency and power extraction for net electricity production with rather conventional technology and/or operational parameters, while an advanced blanket (or several of them) will aim at demonstrating, with limited coverage, features that are deemed necessary for a commercial FPP. Currently, two driver blanket candidates are being investigated for the EU DEMO, namely the water-cooled lithium lead and the helium-cooled pebble bed breeding blanket concepts. The PCD phase has been characterized not only by the detailed design of the BB systems themselves, but also by their holistic integration in DEMO, prioritizing near-term solutions, in accordance with the idea of a driver blanket. This paper summarizes the status for both BB driver blanket candidates at the end of the PCD phase, including their corresponding tritium extraction and removal (TER) systems, underlining the main achievements and lessons learned, exposing outstanding key system design and R&D challenges and presenting identified opportunities to address those risks during the conceptual design (CD) phase that started in 2021. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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12 pages, 4300 KiB

Open AccessArticle

ITER Test Blanket Module—ALARA Investigations for Port Cell Pipe Forest Replacement

by Jean-Pierre Friconneau, Tristan Batal, Olivier David, Chiara Di Paolo, Fabien Ferlay, Stéphane Gazzotti, Luciano Giancarli, Christophe Lacroix, Jean-Pierre Martins, Benjamin Michel and Jean-Marcel Travere

J. Nucl. Eng. 2023, 4(1), 297-308; https://doi.org/10.3390/jne4010022 - 17 Mar 2023

Viewed by 1998

Abstract

The objective of the ITER test blanket module (TBM) program is to provide experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. The ITER test blanket modules are installed and operated inside the vacuum vessel (VV) at [...] Read more.

The objective of the ITER test blanket module (TBM) program is to provide experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. The ITER test blanket modules are installed and operated inside the vacuum vessel (VV) at the equatorial ports located within port plugs (PP), and each PP includes two TBMs. After each 18-month-long plasma operation campaign, the TBM research plan testing program requires the replacement of the TBMs with new ones during the ITER long-term shutdown, called long-term maintenance (LTM). The replacement of a TBM requires the removal/reinstallation of all test blanket system (TBS) equipment present in the port cell (PC), including those in the port interspace (PI), called pipe forest (PF). TBSs shall be designed so that occupational radiation exposure (ORE) can be as low as reasonably achievable (ALARA) over the life of the plant to follow the ITER policy. To implement ALARA process requirements, design activities shall consider careful integration investigations starting from the early phase to address all engineering aspects of the replacement sequence. The case study focuses on the PF replacement, in particular the port cell operations. This paper describes the investigations and findings of the ALARA optimisation process implementation in the early engineering phase of the PF. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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19 pages, 9063 KiB

Open AccessArticle

Heat Pipe-Based DEMO Divertor Target Concept: High Heat Flux Performance Evaluation

by Wen Wen, Bradut-Eugen Ghidersa, Wolfgang Hering, Jörg Starflinger and Robert Stieglitz

J. Nucl. Eng. 2023, 4(1), 278-296; https://doi.org/10.3390/jne4010021 - 9 Mar 2023

Cited by 1 | Viewed by 2070

Abstract

The use of heat pipes (HP) for the DEMO in-vessel plasma-facing components (PFCs) has been considered because of their high capacity to transport the heat from a heat source to a heat sink by means of the vaporization and condensation of the working [...] Read more.

The use of heat pipes (HP) for the DEMO in-vessel plasma-facing components (PFCs) has been considered because of their high capacity to transport the heat from a heat source to a heat sink by means of the vaporization and condensation of the working fluid inside and their ability to enlarge the heat transfer area of the cooling circuit substantially. Recent engineering studies conducted in the framework of the EUROfusion work package Divertor (Wen et al, 2021) indicate that it is possible to design a heat pipe with a capillary limit above 6 kW using a composite capillary structure (wherein axial grooves cover the adiabatic zone and the condenser, and sintered porous material covers the evaporator). This power level would correspond to an applied heat flux of 20 MW/m², rendering such a design interesting with respect to a divertor target concept. To validate the results of the initial engineering analysis, several experiments have been conducted to evaluate the actual performance of the proposed heat pipe concept. The present contribution presents the experiment’s results regarding the examination of the operating limits of two different designs for an evaporator: one featuring a plain porous structure, and one featuring ribs and channels. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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13 pages, 3155 KiB

Open AccessArticle

Concept of Contamination Control Door for DEMO and Proof of Principle Design

by Yan Wang, Jan Oellerich, Carsten Baars and Martin Mittwollen

J. Nucl. Eng. 2023, 4(1), 228-240; https://doi.org/10.3390/jne4010018 - 1 Mar 2023

Cited by 2 | Viewed by 1826

Abstract

During the maintenance period of a future fusion reactor power plant, called DEMOnstration Power Plant (DEMO), remotely handled casks are required to confine and handle DEMO in-vessel components during their transportation between the reactor and the active maintenance facility. In order to limit [...] Read more.

During the maintenance period of a future fusion reactor power plant, called DEMOnstration Power Plant (DEMO), remotely handled casks are required to confine and handle DEMO in-vessel components during their transportation between the reactor and the active maintenance facility. In order to limit the dispersion of activated dust, a Contamination Control Door (CCD) is designed to be placed at an interface between separable containments (e.g., vacuum vessels and casks) to inhibit the release of contamination at the interface between them. The remotely operated CCD—technically, a double lidded door system—consists of two separable doors (the cask door and port door) and three different locking mechanisms: (i) between the cask door and cask, (ii) between the cask door and port door and (iii) between the port door and port. The locking mechanisms are selected and assessed according to different criteria, and the structure of the CCD is optimized using an Abaqus Topology Optimization Module. Due to the elastic properties of the CCD, deflections will occur during the lifting procedure, which may lead to malfunctions of the CCD. A test rig is developed to investigate the performance of high-risk components in the CCD in the case of deflections and also malpositioning. Misalignment can be induced along three axes and three angles intentionally to test the single components and items. The aim is to identify a possible range of operating in the case of misalignments. It is expected that the proposed CCD design should be able to operate appropriately in the case of ±3 mm translational misalignments and ±1° rotational misalignments. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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9 pages, 1896 KiB

Open AccessArticle

Bubble Formation in ITER-Grade Tungsten after Exposure to Stationary D/He Plasma and ELM-like Thermal Shocks

by Mauricio Gago, Arkadi Kreter, Bernhard Unterberg and Marius Wirtz

J. Nucl. Eng. 2023, 4(1), 204-212; https://doi.org/10.3390/jne4010016 - 21 Feb 2023

Cited by 1 | Viewed by 2042

Abstract

Plasma-facing materials (PFMs) in the ITER divertor will be exposed to severe conditions, including exposure to transient heat loads from edge-localized modes (ELMs) and to plasma particles and neutrons. Tungsten is the material chosen as PFM for the ITER divertor. In previous tests, [...] Read more.

Plasma-facing materials (PFMs) in the ITER divertor will be exposed to severe conditions, including exposure to transient heat loads from edge-localized modes (ELMs) and to plasma particles and neutrons. Tungsten is the material chosen as PFM for the ITER divertor. In previous tests, bubble formation in ITER-grade tungsten was detected when exposed to fusion relevant conditions. For this study, ITER-grade tungsten was exposed to simultaneous ELM-like transient heat loads and D/He (6%) plasma in the linear plasma device PSI-2. Bubble formation was then investigated via SEM micrographs and FIB cuts. It was found that for exposure to 100.000 laser pulses of 0.6 GWm⁻² absorbed power density (P_abs), only small bubbles in the nanometer range were formed close to the surface. After increasing P_abs to 0.8 and 1.0 GWm⁻², the size of the bubbles went up to about 1 µm in size and were deeper below the surface. Increasing the plasma fluence had an even larger effect, more than doubling bubble density and increasing bubble size to up to 2 µm in diameter. When using deuterium-only plasma, the samples showed no bubble formation and reduced cracking, showing such bubble formation is caused by exposure to helium plasma. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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12 pages, 10119 KiB

Open AccessCommunication

Investigation of Electromagnetic Sub-Modeling Procedure for the Breeding Blanket System

by Ivan Alessio Maione, Massimo Roccella and Flavio Lucca

J. Nucl. Eng. 2023, 4(1), 165-176; https://doi.org/10.3390/jne4010013 - 30 Jan 2023

Cited by 2 | Viewed by 1537

Abstract

The outcome of the electromagnetic (EM) analyses carried out during the DEMO pre-conceptual phase demonstrated that EM loads are relevant for the structural assessment of the breeding blanket (BB) and, in particular, for the definition of the boundary conditions at the attachment system [...] Read more.

The outcome of the electromagnetic (EM) analyses carried out during the DEMO pre-conceptual phase demonstrated that EM loads are relevant for the structural assessment of the breeding blanket (BB) and, in particular, for the definition of the boundary conditions at the attachment system with the vacuum vessel. However, within the scope of the previous campaign, the results obtained using simplified models only give a rough estimation of the EM loads inside the BB structure. This kind of data has been considered suitable for a preliminary assessment of the BB segments, but it is not considered representative as input for structural analysis in which a detailed BB internal structure (that considers cooling channels, thin plates, etc.) is analyzed. Indeed, mesh dimensions and computational time usually limit EM models that simulate a whole DEMO sector. In many cases, these constraints lead to a strong homogenization of the BB structure, not allowing the calculation of the EM loads on the internal structure with high precision. To overcome such limitations, an EM sub-modeling procedure was investigated using ANSYS EMAG. The sub-modeling feasibility is studied using the rigid boundary condition method. This method consists of running a global “coarse” mesh, including all the conducting structures that can have some impact on the component under investigation and inputting the obtained results on the detailed sub-model of the structure of interest as time-varying boundary conditions. The procedure was tested on the BB internal structure, taking as reference a DEMO 2017 baseline sector and the helium cooled pebble bed (HCPB) concept with its complex internal structure made by pins. The obtained results show that the method is also reliable in the presence of non-linear magnetic behaviour. The methodology is proposed for application in future BB system assessments. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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10 pages, 2059 KiB

Open AccessArticle

Development of the W7-X Alkali Metal Beam Diagnostic Observation System for OP2

by Domonkos Nagy, Sándor Zoletnik, Matthias Otte, Miklós Vécsei, Maciej Krychowiak, Ralf König, Dániel Dunai, Gábor Anda, Sándor Hegedűs, Barnabás Csillag, Imre Katona and W7-X Team

J. Nucl. Eng. 2023, 4(1), 142-151; https://doi.org/10.3390/jne4010010 - 18 Jan 2023

Cited by 2 | Viewed by 1747

Abstract

On a Wendelstein 7-X (W7-X), an alkali metal beam (AMB) diagnostic system was installed in order to measure the plasma edge electron density profiles and turbulence transport. A sodium beam was injected in the plasma, and the light emission was observed by an [...] Read more.

On a Wendelstein 7-X (W7-X), an alkali metal beam (AMB) diagnostic system was installed in order to measure the plasma edge electron density profiles and turbulence transport. A sodium beam was injected in the plasma, and the light emission was observed by an optical system. During the last operation phase, OP1.2b campaign trial spectral measurements were performed with a dedicated optical branch. The results showed the emergence of potential CX lines in the light spectra during sodium injection. The lines were identified as Carbon III, which were the dominant lines observed by other diagnostics at the edge plasma. Based on these results, an additional dedicated optical system was developed and installed in 2021 for the upcoming operational phase, OP2. The optics were designed for multiple purposes: spectral measurements for the AMB system and for a He/Ne gas jet. The system was designed to allow implementation of further diagnostics on this port later (e.g., coherence imaging system). The details of the implementation of the design requirements and the main challenges of the manufacturing process and installation are discussed in this paper. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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16 pages, 4831 KiB

Open AccessArticle

Development of Mechanical Pipe-Connection Design for DEMO

by Viktor Milushev, Azman Azka and Martin Mittwollen

J. Nucl. Eng. 2023, 4(1), 111-126; https://doi.org/10.3390/jne4010008 - 11 Jan 2023

Cited by 2 | Viewed by 1776

Abstract

Maintenance of the DEMO breeding blanket includes the removal and replacement of plasma-facing components. To access the breeding blanket, multiple coolant pipes need to be removed to allow access to the tokamak. As an option to reduce downtime and increase maintenance speed, the [...] Read more.

Maintenance of the DEMO breeding blanket includes the removal and replacement of plasma-facing components. To access the breeding blanket, multiple coolant pipes need to be removed to allow access to the tokamak. As an option to reduce downtime and increase maintenance speed, the pipe-connection concept is developed to allow the removal of multiple pipes at the same time using a remotely operated mechanical connection. The remotely operated multi-pipe Mechanical Pipe Connection (MPC) needs to fulfil multiple requirements, such as high operating temperature and high external forces while at the same time maintaining an acceptable level of sealing between the high-pressure fluid and vacuum surroundings. In addition to the external conditions, the pipes of multiple sizes and fluids are connected in a manifold configuration. Although this will reduce the overall time required to operate the mechanical pipe connection when compared to multiple single-pipe connections, this will introduce additional forces and stresses due the interaction between pipe flow (e.g., simultaneous high- and low-temperature fluid pipes on the same manifold) through the manifold flange. The requirements and the boundary conditions of the multi-pipe MPC are taken into consideration during the design process of MPC. The design process is carried out to find the optimum form and size to allow the mechanical function of the pipe connection during the maintenance phase while withstanding the extreme operating conditions that the MPC will face the during operational phase. The resulting design will then be analyzed using numerical methods to assess the capability of the MPC designs. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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15 pages, 5009 KiB

Open AccessArticle

Ex Situ LIBS Analysis of WEST Divertor Wall Tiles after C3 Campaign

by Indrek Jõgi, Peeter Paris, Elodie Bernard, Mathilde Diez, Emmanuelle Tsitrone, Antti Hakola, Jari Likonen, Tomi Vuoriheimo, Eduard Grigore, the WEST Team and EUROfusion WP PFC/PWIE Contributors

J. Nucl. Eng. 2023, 4(1), 96-110; https://doi.org/10.3390/jne4010007 - 5 Jan 2023

Cited by 1 | Viewed by 1697

Abstract

Fuel retention monitoring in tokamak walls requires the development of remote composition analysis methods such as laser-induced breakdown spectroscopy (LIBS). The present study investigates the feasibility of the LIBS method to analyse the composition and fuel retention in three samples from WEST divertor [...] Read more.

Fuel retention monitoring in tokamak walls requires the development of remote composition analysis methods such as laser-induced breakdown spectroscopy (LIBS). The present study investigates the feasibility of the LIBS method to analyse the composition and fuel retention in three samples from WEST divertor erosion marker tiles after the experimental campaign C3. The investigated samples originated from tile regions outside of strong erosion and deposition regions, where the variation of thin deposit layers is relatively small and facilitates cross-comparison between different analysis methods. The depth profiles of main constituents W, Mo and C were consistent with depth profiles determined by other composition analysis methods, such as glow-discharge optical emission spectroscopy (GDOES) and secondary ion mass spectrometry (SIMS). The average LIBS depth resolution determined from depth profiles was 100 nm/shot. The averaging of the spectra collected from multiple spots of a same sample allowed us to improve the signal-to-noise ratio, investigate the presence of fuel D and trace impurities such as O and B. In the investigated tile regions with negligible erosion and deposition, these impurities were clearly detectable during the first laser shot, while the signal decreased to noise level after a few subsequent laser shots at the same spot. LIBS investigation of samples originating from the deposition regions of tiles may further clarify LIBS’ ability to investigate trace impurities. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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10 pages, 11377 KiB

Open AccessArticle

Integrated Design of the Vacuum and Safety Barrier between the Lithium and Test Systems of IFMIF-DONES

by András Zsákai, Tamás Dézsi, András Korossy-Khayll, Imre Katona, Viktor Varga, Endre Kósa, Dénes Zoltán Oravecz, Santiago Becerril, Carlos Meléndez, Jesus Castellanos, Gioacchino Micciché and Angel Ibarra

J. Nucl. Eng. 2023, 4(1), 49-58; https://doi.org/10.3390/jne4010004 - 26 Dec 2022

Cited by 2 | Viewed by 1625

Abstract

The international fusion materials irradiation facility-DEMO-oriented neutron source (IFMIF-DONES) is a facility that is designed under the framework of the EU fusion roadmap. It is going to be an essential irradiation facility for testing and qualifying candidate materials under severe irradiation conditions of [...] Read more.

The international fusion materials irradiation facility-DEMO-oriented neutron source (IFMIF-DONES) is a facility that is designed under the framework of the EU fusion roadmap. It is going to be an essential irradiation facility for testing and qualifying candidate materials under severe irradiation conditions of a neutron field having an energy spectrum like the one present in a fusion power reactor. The material specimens are irradiated in a containment structure named the test cell (TC), which is part of the test systems (TS). The TC also houses a part of the other major system (lithium system, LS), which provides the liquid lithium for the reaction through a piping system. At a point, the lithium piping needs to exit the TS, but the primary safety boundary must be continuous around these penetrations. Therefore, a special barrier, called the test systems–lithium systems interface cell (TLIC), has been developed around the piping system to provide a safety-approved and remotely maintainable vacuum boundary envelope. In this paper, the integrated design development of the TLIC is described, consisting of the design development according to the RCC-MRx code, the remote-handling (RH) needs, and the procedures and safety-related special needs of the design. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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21 pages, 7539 KiB

Open AccessArticle

Status of Scoping Nuclear Analyses for the Evolving Design of ITER TBM Port Cells

by Moataz Harb, Dieter Leichtle, Byoung-Yoon Kim, Jean-Pierre Martins, Eduard Polunovskiy, Jayant Somvanshi and Jaap G. van der Laan

J. Nucl. Eng. 2023, 4(1), 28-48; https://doi.org/10.3390/jne4010003 - 23 Dec 2022

Viewed by 1792

Abstract

ITER is an international collaborative effort towards the realization of fusion energy via the magnetic confinement concept. Two of the equatorial ports in the facility are dedicated to the testing of tritium breeding concepts, which is essential for the tritium self-sufficiency of future [...] Read more.

ITER is an international collaborative effort towards the realization of fusion energy via the magnetic confinement concept. Two of the equatorial ports in the facility are dedicated to the testing of tritium breeding concepts, which is essential for the tritium self-sufficiency of future fusion reactors. The concerned Test Blanket System (TBS) consists of a Test Blanket Module (TBM) residing inside the TBM–Port Plug (TBM-PP) and its associated ancillary systems in the Tokamak facility. In this paper, the results of a full suite of nuclear analyses concerning the shielding performance of the Pipe Forest (PF) and Bioshield Plug (BP), to reflect on the evolution of their designs, are discussed. On the BP side, the design of the peripheral part has been reviewed considering the ventilation openings and butterfly doors, to assure the design compliance with the Radiation Map (RadMap) requirements for the neutron flux in the Port Cell (PC), behind the BP. On the PF side, the pipes routing and maintenance corridor door have been redesigned, by taking into account results from previously concluded nuclear analyses. The neutronics model was developed from CAD and was used to perform transport simulations in two plasma modes: on and off. For plasma-on mode, the plasma neutron field in the Port Interspace (PI) as well as behind the BP was assessed and few shielding options were explored. The responses due to decay neutrons from ¹⁷N in activated cooling water were also considered. For the plasma-off mode, the focus was shifted to further refine the ShutDown Dose Rate (SDDR) maps, which is of importance for maintenance operations that are foreseen to take place at various stages of ITER operation, in particular following the FPO-1, FPO-2, and Short operation scenarios. In addition, detailed activation analyses were carried out to provide a provisional waste classification. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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17 pages, 3458 KiB

Open AccessArticle

Experimental Thermal–Hydraulic Testing of a Mock-Up of the Fuel-Breeder Pin Concept for the EU-DEMO HCPB Breeding Blanket

by Ali Abou-Sena, Bradut-Eugen Ghidersa, Guangming Zhou, Joerg Rey, Francisco A. Hernández, Martin Lux and Georg Schlindwein

J. Nucl. Eng. 2023, 4(1), 11-27; https://doi.org/10.3390/jne4010002 - 22 Dec 2022

Cited by 1 | Viewed by 1968

Abstract

The fusion program in the Karlsruhe Institute of Technology (KIT) leads the R&D of the DEMO helium-cooled pebble bed (HCPB) breeding blanket within the work package breeding blanket (WPBB) of the Eurofusion Consortium in the European Union (EU). A new design of the [...] Read more.

The fusion program in the Karlsruhe Institute of Technology (KIT) leads the R&D of the DEMO helium-cooled pebble bed (HCPB) breeding blanket within the work package breeding blanket (WPBB) of the Eurofusion Consortium in the European Union (EU). A new design of the HCPB breeder zone, with a layout inspired by a nuclear reactor fuel rod arrangement, was developed recently and called the fuel-breeder pin concept. In addition, a mock-up (MU) of this fuel-breeder pin was designed and manufactured at KIT in order to test and validate its thermal–hydraulic performance. This paper reports on the results of the first experimental campaign dedicated to the fuel-breeder pin MU testing that was performed in the Helium Loop Karlsruhe (HELOKA) facility. The paper presents: (i) the integration of the fuel-breeder pin MU into the HELOKA loop including considerations of the experimental set-up, (ii) an overview of the plan for the experimental campaigns, and (iii) a discussion of the experimental results with a focus on aspects relevant for the validation of the thermal–hydraulic design of the HCPB breeder zone. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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10 pages, 3090 KiB

Open AccessArticle

Development and Basic Qualification Steps towards an Electrochemically Based H-Sensor for Lithium System Applications

by Nils Holstein, Wolfgang Krauss and Francesco Saverio Nitti

J. Nucl. Eng. 2023, 4(1), 1-10; https://doi.org/10.3390/jne4010001 - 21 Dec 2022

Cited by 1 | Viewed by 1495

Abstract

IFMIF-DONES, or the InternationalFusionMaterialsIrradiationFacility-DEMOOrientedNeutronSource, is a facility for investigations into foreseen fusion power plant materials using the relevant neutron irradiation of 14 MeV. This [...] Read more.

IFMIF-DONES, or the InternationalFusionMaterialsIrradiationFacility-DEMOOrientedNeutronSource, is a facility for investigations into foreseen fusion power plant materials using the relevant neutron irradiation of 14 MeV. This special n-irradiation is generated by the interaction of deuteron beams with liquid lithium. A critical issue during the operation of IFMIF-DONES is the enrichment of dissolved impurities in the Li-melt loops. The danger occurs as a result of hydrogen-induced corrosivity and embrittlement of the loop components, as well as the security hazards associated with the radioactive tritium. Hence, the application of liquid lithium in IFMIF-DONES requires a suitable impurity control system for reliable and low-level maintenance under the operating conditions of DONES. Regarding those requirements, an electrochemical sensor for hydrogen monitoring was developed in the frame of an international EUROFusion–WPENS task, to determine H-concentrations via the electro-motive force (EMF) of Li-melts and a suitable online-monitoring system. Long-term tests demonstrated that the sensor fulfills the requirements of chemical and mechanical stability and functionality under the harsh Li environment under the planned DONES conditions. Obtained results and operational experiences will be discussed in regard to application windows, reproducibility and calibration needs. Additionally, recommendations will be outlined for upgraded systems and future qualification needs. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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7 pages, 1346 KiB

Open AccessArticle

A Deep Learning-Based Method to Detect Hot-Spots in the Visible Video Diagnostics of Wendelstein 7-X

by Máté Szűcs, Tamás Szepesi, Christoph Biedermann, Gábor Cseh, Marcin Jakubowski, Gábor Kocsis, Ralf König, Marco Krause, Aleix Puig Sitjes and the W7-X Team

J. Nucl. Eng. 2022, 3(4), 473-479; https://doi.org/10.3390/jne3040033 - 15 Dec 2022

Cited by 2 | Viewed by 1681

Abstract

Wendelstein 7-X (W7-X) is currently the largest optimized stellarator in operation in the world. Its main objective is to demonstrate long pulse operation and to investigate the suitability of this type of fusion device for a power plant. Maintaining the safety of the [...] Read more.

Wendelstein 7-X (W7-X) is currently the largest optimized stellarator in operation in the world. Its main objective is to demonstrate long pulse operation and to investigate the suitability of this type of fusion device for a power plant. Maintaining the safety of the first wall is critical to achieving the desired discharge times of approximately 30 min while keeping a steady-state condition. We present a deep learning-based solution to detect the unexpected plasma-wall and plasma-object interactions, so-called hot-spots, in the images of the Event Detection Intelligent Camera (EDICAM) system. These events can pose a serious threat to the safety of the first wall, therefore, to the operation of the device. We show that sufficiently training a neural network with relatively small amounts of data is possible using our approach of mixing the experimental dataset with new images containing so-called synthetic hot-spots generated by us. Diversifying the dataset with synthetic hot-spots increases performance and can make up for the lack of data. The best performing YOLOv5 Small model processes images in 168 ms on average during inference, making it a good candidate for real-time operation. To our knowledge, we are the first ones to be able to detect events in the visible spectrum in stellarators with high accuracy, using neural networks trained on small amounts of data while achieving near-real-time inference times. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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12 pages, 4633 KiB

Open AccessArticle

Design Features and Simulation of the New-Build HELOKA-US Facility for the Validation of the DEMO Helium-Cooled Pebble Bed Intermediate Heat Transport and Storage System

by Xiaoyang Gaus-Liu, Evaldas Bubelis, Sara Perez-Martin, Bradut-Eugen Ghidersa and Wolfgang Hering

J. Nucl. Eng. 2022, 3(4), 461-472; https://doi.org/10.3390/jne3040032 - 14 Dec 2022

Viewed by 1610

Abstract

For the EU-DEMO Helium-Cooled Pebble Bed (HCPB) concept, an indirect coupled design (ICD) with a molten salt (MS) loop as an intermediate heat transport and storage system (IHTS) is considered for the conceptual design phase. The IHTS with an energy storage decouples the [...] Read more.

For the EU-DEMO Helium-Cooled Pebble Bed (HCPB) concept, an indirect coupled design (ICD) with a molten salt (MS) loop as an intermediate heat transport and storage system (IHTS) is considered for the conceptual design phase. The IHTS with an energy storage decouples the primary heat transport system (PHTS) that undergoes pulse and dwell power cycles from the power conversion system (PCS), and thus can provide stable power to the turbine and grid. However, the maintenance of stable He and MS parameters during transitions from dwell to pulse and vice versa is challenging for the design of the MS loop, and the real performance of the helium–MS heat exchanger (He/MS HX) shall be verified. To investigate such components and conditions, a new R&D infrastructure HELOKA-US (Helium Loop Karlsruhe—Upgrade Storage) is under construction for the validation of prototypical components and the MS loop operation under stationary and transitional conditions. This paper provides the design features of Phase 1a of the project and the simulation results with EBSILON on the power generation phase. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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8 pages, 1247 KiB

Open AccessArticle

Titanium and Tantalum Used as Functional Gradient Interlayer to Join Tungsten and Eurofer97

by Marianne Richou, Isabelle Chu, Geoffrey Darut, Raphael Maestracci, Manda Ramaniraka and Erick Meillot

J. Nucl. Eng. 2022, 3(4), 453-460; https://doi.org/10.3390/jne3040031 - 13 Dec 2022

Cited by 1 | Viewed by 1558

Abstract

For the DEMO reactor, tungsten is considered as an armor material. Eurofer97 is planned to be used as a structural material for the first wall and in the divertor region, especially for the shielding liner component. To date, several joining solutions between W [...] Read more.

For the DEMO reactor, tungsten is considered as an armor material. Eurofer97 is planned to be used as a structural material for the first wall and in the divertor region, especially for the shielding liner component. To date, several joining solutions between W and Eurofer97 have been developed (copper brazing, W and Eurofer97 functional gradient material (FGM), etc.). Each existing joining solution has its own advantages (joining material, improved manufacturing process). In the present study, the choice of the joining material is driven, among other constraints, by a desire to minimize the thermal stresses at the materials’ interface. In this regard, FGM represents a promising solution. Another constraint that is taken into account in this study concerns the manufacturing process involved, which should be an improved industrial process. The present study proposes a joining solution, based on FGM, which, additionally to the advantages of the existing solutions, could reduce the long-term activation of the joining material. The development of a joining solution via Ti and Ta as materials constituting the FGM (Ti/Ta FGM) is presented in this paper. Due to the achieved density and the composition’s accuracy, the cold spray process is shown to be adapted for the Ti/Ta FGM’s manufacturing. Based on the feedback on the experience of joining between W, W/Cu FGM and CuCrZr, the final joining between W, Ti/Ta FGM and Eurofer97 is achieved using hot isostatic pressing, followed by a thermal treatment to recover Eurofer97’s mechanical properties, resulting in good joining quality. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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7 pages, 3259 KiB

Open AccessArticle

Powder Metallurgy Produced Aligned Long Tungsten Fiber Reinforced Tungsten Composites

by Yiran Mao, Jan W. Coenen, Chao Liu, Alexis Terra, Xiaoyue Tan, Johann Riesch, Till Höschen, Yucheng Wu, Christoph Broeckmann and Christian Linsmeier

J. Nucl. Eng. 2022, 3(4), 446-452; https://doi.org/10.3390/jne3040030 - 8 Dec 2022

Cited by 8 | Viewed by 1760

Abstract

For the future fusion reactor, tungsten is the main candidate material as the plasma-facing material. However, considering the high thermal stress during operation, the intrinsic brittleness of tungsten is one of the issues. To overcome the brittleness, tungsten fiber reinforces tungsten composites (W [...] Read more.

For the future fusion reactor, tungsten is the main candidate material as the plasma-facing material. However, considering the high thermal stress during operation, the intrinsic brittleness of tungsten is one of the issues. To overcome the brittleness, tungsten fiber reinforces tungsten composites (W_f/W) developed using extrinsic toughening mechanisms. The powder metallurgy process and chemical vapor deposition process are the two production routes for preparing W_f/W. For the powder metallurgy route, due to technical limitations, previous studies focused on short random distributed fiber-reinforced composites. However, for short random fiber composites, the strength and reinforcement effect are considerably limited compared to aligned continuous fiber composites. In this work, aligned long tungsten fiber reinforced tungsten composites have been first time realized based on powder metallurgy processes, by alternately placing tungsten weaves and tungsten powder layers. The produced W_f/W shows significantly improved mechanical properties compared to pure W and conventional short fiber W_f/W. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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11 pages, 1256 KiB

Open AccessArticle

R&D Needs for the Design of the EU-DEMO HCPB ICD Balance of Plant in FP9

by Sara Perez-Martin, Evaldas Bubelis, Wolfgang Hering and Luciana Barucca

J. Nucl. Eng. 2022, 3(4), 435-445; https://doi.org/10.3390/jne3040029 - 6 Dec 2022

Viewed by 1651

Abstract

During the Pre-Conceptual Design Phase of the EU-DEMO, two BOP solutions for WCLL and HCPB were elaborated, as close as possible to industrial standards. Nevertheless, each solution has open issues to be investigated, analytically and experimentally, in the Conceptual Design Phase (CDP). For [...] Read more.

During the Pre-Conceptual Design Phase of the EU-DEMO, two BOP solutions for WCLL and HCPB were elaborated, as close as possible to industrial standards. Nevertheless, each solution has open issues to be investigated, analytically and experimentally, in the Conceptual Design Phase (CDP). For the HCPB, the functionality and operability of the Helium-Molten Salt Heat Exchanger, and the coupling to a helium loop with a prototypic helium blower, is of primary interest. In addition, the operation of the pulse, dwell and transitions will be investigated within the new build infrastructure, HELOKA-US (Upgrade Storage), to be erected at KIT. The design requires a certain flexibility, since the final parameters of the Primary Heat Transfer System of DEMO may vary, due to plasma optimizations during CDP. HELOKA-US benefits from the high-pressure helium loop HELOKA-HP, erected to test HCPB-Breeding Blanket and First Wall modules, as well as from the competencies of preparing, handling and testing of various molten salts used for heat transfer optimization and natural convection. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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14 pages, 26491 KiB

Open AccessArticle

Thermomechanical Analysis of a PFC Integrating W Lattice Armour in Response to Different Plasma Scenarios Predicted in the EU-DEMO Tokamak

by Damiano Paoletti, Pierluigi Fanelli, Riccardo De Luca, Chiara Stefanini, Francesco Vivio, Valerio Gioachino Belardi, Simone Trupiano, Giuseppe Calabrò, Jeong-Ha You and Rudolf Neu

J. Nucl. Eng. 2022, 3(4), 421-434; https://doi.org/10.3390/jne3040028 - 2 Dec 2022

Cited by 2 | Viewed by 1664

Abstract

Despite the high performance exhibited by tungsten (W), no material would be able to withstand the huge loads expected with extreme plasma transients in EU-DEMO and future reactors, where the installation of sacrificial first wall limiters is essential to prevent excessive wall degradation. [...] Read more.

Despite the high performance exhibited by tungsten (W), no material would be able to withstand the huge loads expected with extreme plasma transients in EU-DEMO and future reactors, where the installation of sacrificial first wall limiters is essential to prevent excessive wall degradation. The integration of W lattices in the architecture of such components can allow for meeting their conflictual requirements: indeed, they must ensure the effective exhaust of the nominal thermal load during stationary operation; when transients occur, they must thermally insulate and decouple the surface from the heat sink, promoting prompt vapour shielding formation. Starting from the optimised layouts highlighted in a previous study, in this work, a detailed 3D finite element model was developed to analyse in depth the influence of the actual features of the latticed metamaterial on the overall performance of the EU-DEMO limiter PFC on the basis of a flat tile configuration. Its main goal is to help in identifying the most promising layout as a preconceptual design for the fabrication of a small-scale mock-up. For this purpose, the complex geometry of a W-based lattice armour was faithfully reproduced in the model and analysed. This allowed for a detailed assessment of the thermally induced stresses that develop in the component because of the temperature field in response to a number of plasma scenarios—above all, normal operation and ramp down. Structural integrity was verified through the acceptance criteria established for ITER. The two optimised layouts proposed for the PFC were able to effectively meet the requirements under normal reactor operating conditions, while they missed some requirements in the ramp-down case. However, the first HHF tests will be performed in order to benchmark the analyses. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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12 pages, 6907 KiB

Open AccessArticle

CFD Analysis and Optimization of the DEMO WCLL Central Outboard Segment Bottom-Cap Elementary Cell

by Lorenzo Melchiorri, Pietro Arena, Fabio Giannetti, Simone Siriano and Alessandro Tassone

J. Nucl. Eng. 2022, 3(4), 409-420; https://doi.org/10.3390/jne3040027 - 1 Dec 2022

Cited by 2 | Viewed by 1509

Abstract

In the design of magnetic confinement nuclear fusion power plants, the breeding blanket (BB) plays a crucial role since it must fulfil key functions such as tritium breeding, radiation-shielding, and removal of the heat power generated by the plasma. The latter task is [...] Read more.

In the design of magnetic confinement nuclear fusion power plants, the breeding blanket (BB) plays a crucial role since it must fulfil key functions such as tritium breeding, radiation-shielding, and removal of the heat power generated by the plasma. The latter task is achieved by the first wall (FW) and breeding zone (BZ) cooling systems, which in the water-cooled lithium–lead (WCLL) BB employ pressurized water. Different arrangements of BZ coolant conduits have been investigated in the recent past to identify an efficient layout, which could meet the structural materials’ operational temperature constraint and which could provide the optimal coolant outlet temperature. However, most of the computational fluid dynamic (CFD) analyses that have been carried out until now have been focused on the equatorial WCLL elementary cell of the central outboard segment (COB). The aim of this work is to broaden the analysis to other relevant locations in the blanket. An assessment of the design of the cooling system of the COB bottom-cap elementary BZ cell has been identified as a top design priority due to its different geometry and thermal loads. The cooling efficiency of the BZ and FW systems is investigated to assess if the coolant-appropriate design conditions are matched and the temperature distribution in the cell is analyzed to identify the onset of hot spots. Different layouts of the FW systems are proposed and compared in terms of thermal–hydraulic reliability. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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11 pages, 5221 KiB

Open AccessArticle

Swelling of Highly Neutron Irradiated Beryllium and Titanium Beryllide

by Vladimir Chakin, Alexander Fedorov, Ramil Gaisin and Milan Zmitko

J. Nucl. Eng. 2022, 3(4), 398-408; https://doi.org/10.3390/jne3040026 - 28 Nov 2022

Cited by 6 | Viewed by 1727

Abstract

The swelling of beryllium and titanium beryllide after irradiation at 70–750 °C to neutron fluences of (0.25–8) · 10²² cm⁻² (E > 1 MeV) was measured using methods of immersion, dimension, and helium pycnometry. Dependences of the swelling on the irradiation [...] Read more.

The swelling of beryllium and titanium beryllide after irradiation at 70–750 °C to neutron fluences of (0.25–8) · 10²² cm⁻² (E > 1 MeV) was measured using methods of immersion, dimension, and helium pycnometry. Dependences of the swelling on the irradiation temperature and neutron dose were plotted and analyzed. The dose dependences show linear dependences of the swelling for all irradiation temperatures except 70 °C, where the swelling rate varies, depending on increasing neutron dose. Be-7Ti shows much less swelling than pure Be. Irradiation at 430–750 °C to neutron fluence of 1.82 · 10²² cm⁻² (E > 1 MeV) leads to swelling of Be at about 50%; for Be-7Ti, it is 2.7%. The microstructure study shows that the formation of bubbles and pores in beryllium occurs much more intense than in titanium beryllide. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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13 pages, 15807 KiB

Open AccessArticle

Potential Use of IFMIF-DONES Target Back-Plate for Material Specimens

by Yuefeng Qiu, Frederik Arbeiter, Davide Bernardi, Manuela Frisoni, Sergej Gordeev, Rebeca Hernández and Arkady Serikov

J. Nucl. Eng. 2022, 3(4), 385-397; https://doi.org/10.3390/jne3040025 - 25 Nov 2022

Cited by 3 | Viewed by 1605

Abstract

In the IFMIF-DONES facility of the future, the back-plate behind the Li target will receive strong irradiation from high-energy neutrons. The potential use of the back-plate for material specimens is attractive with respect to providing complementary irradiation data for Eurofer. In this work, [...] Read more.

In the IFMIF-DONES facility of the future, the back-plate behind the Li target will receive strong irradiation from high-energy neutrons. The potential use of the back-plate for material specimens is attractive with respect to providing complementary irradiation data for Eurofer. In this work, DPA (displacement per atom) and gas production rates as well as DPA gradients and temperature distributions have been studied for the center segment of the back-plate, using both a nominal beam and a reduced beam footprint. It is shown that specimens can be produced with high DPA in similar conditions to the DEMO first-wall. Based on the size of the SSTT (small specimen test technology) specimens, the limited number of samples obtainable from the adopted arrangement scheme is driven by a major constraint: the thickness of the back-plate. A parametric study of the back-plate’s thickness provides an alternative arrangement scheme; thus, the DPA and gradient of the specimens are remarkably improved. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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12 pages, 1863 KiB

Open AccessArticle

Implementation of Safety Aspects in IFMIF-DONES Design

by Francisco Martín-Fuertes, Juan Carlos Marugán, Antonio García, Tonio Pinna, Yuefeng Qiu, Atte Helminen, Slawomir Potempski, Eduardo Gallego, Francisco Ogando, Gianluca D’Ovidio, Manuel Pérez and Ángel Ibarra

J. Nucl. Eng. 2022, 3(4), 373-384; https://doi.org/10.3390/jne3040024 - 18 Nov 2022

Cited by 2 | Viewed by 1607

Abstract

Integration of safety aspects in IFMIF-DONES design is a main objective of EUROfusion and European Commission projects. IFMIF-DONES will be a radioactive facility of the first category, and stringent safety objectives must be achieved and demonstrated. A very low acceptable risk for the [...] Read more.

Integration of safety aspects in IFMIF-DONES design is a main objective of EUROfusion and European Commission projects. IFMIF-DONES will be a radioactive facility of the first category, and stringent safety objectives must be achieved and demonstrated. A very low acceptable risk for the worker, the public and the environment is the main principle in the design phase. The progress of safety activities is performed iteratively as detailed engineering develops, taking into account the uniqueness of the facility: a high-power deuterons accelerator (125 mA, 40 MeV), a target of flowing liquid lithium, traps for activation products, a dedicated-design module for irradiated samples, a massive shielding cooled room with confinement function, and a number of conventional systems with safety functions. Several phases are developed: (i) identification of sources and materials at risk, radioactive and nonradioactive, subject to potential mobilization, (ii) failure mode analysis and effects of systems, starting at the functional level, and support with probabilistic analysis, (iii) identification of scenarios leading to unacceptable risk if unmitigated, (iv) proposal of layers of defense by means of safety-credited components and design features, (v) deterministic analysis of scenarios in support of requirements, and (vi) definition and demonstration of safety requirements charged to components. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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9 pages, 2425 KiB

Open AccessArticle

Long Range Optical Distance Sensors for Liquid Metal Free Surface Detection

by Björn Brenneis

J. Nucl. Eng. 2022, 3(4), 364-372; https://doi.org/10.3390/jne3040023 - 16 Nov 2022

Cited by 1 | Viewed by 1604

Abstract

For the demonstration of fusion power plant technology, DEMO dedicated materials are necessary to cope with the harsh environment of high energy neutrons. For this purpose, the international neutron irradiation facility for fusion materials IFMIF/DEMO Oriented Neutron Source (DONES) is planned to be [...] Read more.

For the demonstration of fusion power plant technology, DEMO dedicated materials are necessary to cope with the harsh environment of high energy neutrons. For this purpose, the international neutron irradiation facility for fusion materials IFMIF/DEMO Oriented Neutron Source (DONES) is planned to be built in Granda, Spain. In the DONES facility, a deuteron beam hitting the lithium target produces a high energy neutron flux. Due to the high-power density, the windowless target is a free surface liquid lithium flow in a duct with a concave backplate. In order to keep the heat released by the beam within the liquid lithium and to avoid its intrusion in the backplate, a stable configuration of the free surface flow with a setpoint layer thickness of 25 ± 1 mm is crucial. In particular, stable wave structures, so called wakes, which occur from accumulated impurities at the nozzle edge, can cause a critical local decrease in the layer thickness of more than 1 mm. Therefore, it is necessary to better understand the nature of these wakes and to be able to monitor the surface profile to shut down the beam in case of a critical thickness loss, but to avoid unintended shutdowns. In the context of this work, currently available optical sensors were tested on their capability of detecting a specular liquid metal surface at measurement distances of several meters. After an initial selection, two optical sensors were further considered. Experiments with the liquid metal alloy GaInSn and simulations with the software Blender of the selected optical sensors for their capability of measuring distances of liquid metal were conducted. The results showed a significant dependency of the measurement results on the waviness of the liquid metal surface. Nevertheless, it was possible to resolve the wavy liquid metal surface with a sufficient resolution to detect critical wake structures. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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10 pages, 3759 KiB

Open AccessArticle

The Double-Disk Diamond Window as Backup Broadband Window Solution for the DEMO Electron Cyclotron System

by Gaetano Aiello, Gerd Gantenbein, John Jelonnek, Andreas Meier, Theo Scherer, Sabine Schreck, Dirk Strauss and Manfred Thumm

J. Nucl. Eng. 2022, 3(4), 342-351; https://doi.org/10.3390/jne3040021 - 15 Nov 2022

Cited by 1 | Viewed by 1884

Abstract

The second variant of the electron cyclotron heating and current drive system in DEMO considers the deployment of 2 MW power Gaussian microwave beams to the plasma by frequency steering. Broadband optical grade chemical vapor deposition diamond windows are thus required. The Brewster-angle [...] Read more.

The second variant of the electron cyclotron heating and current drive system in DEMO considers the deployment of 2 MW power Gaussian microwave beams to the plasma by frequency steering. Broadband optical grade chemical vapor deposition diamond windows are thus required. The Brewster-angle window represents the primary choice. However, in the case of showstoppers, the double-disk window is the backup solution. This window concept was used at ASDEX Upgrade for injection of up to 1 MW at four frequencies between 105 and 140 GHz. This paper shows computational fluid dynamics conjugated heat transfer and structural analyses of such a circumferentially water-cooled window design aiming to check whether it might be used for DEMO microwave beam scenarios. This design was then characterized with respect to different parameters. Temperature and thermal stress results showed that it is a feasible window solution for DEMO, but safety margins against limits shall be increased by introducing design features able to make the fluid more turbulent. A first design change is proposed, showing that, in combination with a higher inlet flow rate, the maximum temperature in the disks can be reduced from 238 to 186 °C, leading, therefore, to lower thermal gradients and stresses in the window. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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9 pages, 4237 KiB

Open AccessArticle

Progress in the Realization of µ-Brush W for Plasma-Facing Components

by Daniel Dorow-Gerspach, Thomas Derra, Marius Gipperich, Thorsten Loewenhoff, Gerald Pintsuk, Alexis Terra, Thomas Weber, Marius Wirtz and Christian Linsmeier

J. Nucl. Eng. 2022, 3(4), 333-341; https://doi.org/10.3390/jne3040020 - 8 Nov 2022

Cited by 1 | Viewed by 1639

Abstract

During the service life of plasma-facing components, they are exposed to cyclic stationary and transient thermal loads. The former causes thermal fatigue and potentially detachment between the plasma-facing material tungsten and the structural Cu-based materials (divertor) and steel (first wall). The latter causes [...] Read more.

During the service life of plasma-facing components, they are exposed to cyclic stationary and transient thermal loads. The former causes thermal fatigue and potentially detachment between the plasma-facing material tungsten and the structural Cu-based materials (divertor) and steel (first wall). The latter causes surface roughening, cracking, or even melting, which could drastically increase the erosion rate. Employing thin flexible W wires (W_w) with a diameter of a few hundred µm can reduce mechanical stresses, and we demonstrated their crack resilience against transient loads within first proof of principle studies. Here, status and future paths towards the large-scale production of such W_w assemblies, including techniques for realizing feasible joints with Cu, steel, or W, are presented. Using wire-based laser metal deposition, we were able to create a homogeneous and shallow infiltration of about 200 µm of the W_w assembly with steel. A high-heat-flux test on such a µ-brush (10 × 10 × 5 mm³ W_w on a ~0.5 mm thick steel layer) using 5 MW/m² for 2000 cycles was performed without loss of any wire. Microstructural examination after and infrared analysis during the test showed no significant signs of degradation of the joint. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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15 pages, 17724 KiB

Open AccessArticle

Large-Scale Tungsten Fibre-Reinforced Tungsten and Its Mechanical Properties

by Daniel Schwalenberg, Jan Willem Coenen, Johann Riesch, Till Hoeschen, Yiran Mao, Alexander Lau, Hanns Gietl, Leonard Raumann, Philipp Huber, Christian Linsmeier and Rudolf Neu

J. Nucl. Eng. 2022, 3(4), 306-320; https://doi.org/10.3390/jne3040018 - 3 Nov 2022

Cited by 8 | Viewed by 2011

Abstract

Tungsten-fibre-reinforced tungsten composites (

W_{f} / W

) have been in development to overcome the inherent brittleness of tungsten as one of the most promising candidates for the first wall and divertor armour material in a future fusion power plant. As the [...] Read more.

Tungsten-fibre-reinforced tungsten composites (

W_{f} / W

) have been in development to overcome the inherent brittleness of tungsten as one of the most promising candidates for the first wall and divertor armour material in a future fusion power plant. As the development of

W_{f} / W

continues, the fracture toughness of the composite is one of the main design drivers. In this contribution, the efforts on size upscaling of

W_{f} / W

based on Chemical Vapour Deposition (CVD) are shown together with fracture mechanical tests of two different size samples of

W_{f} / W

produced by CVD. Three-point bending tests according to American Society for Testing and Materials (ASTM) Norm E399 for brittle materials were used to obtain a first estimation of the toughness. A provisional fracture toughness value of up to

346 MPa m^{1 / 2}

was calculated for the as-fabricated material. As the material does not show a brittle fracture in the as-fabricated state, the J-Integral approach based on the ASTM E1820 was additionally applied. A maximum value of the J-integral of

41 {kJ / m}^{2}

(

134.8 MPa m^{1 / 2}

) was determined for the largest samples. Post mortem investigations were employed to detail the active mechanisms and crack propagation. Full article

(This article belongs to the Special Issue Special Issue Dedicated to 32nd Symposium on Fusion Technology—SOFT2022)

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