Fusion Engineering and Design最新文献

{"title":"Enhancing computational efficiency in nuclear fusion through reduced order modelling: Applications in magnetohydrodynamics","authors":"Matteo Lo Verso ,&nbsp;Stefano Riva ,&nbsp;Carolina Introini ,&nbsp;Eric Cervi ,&nbsp;Luciana Barucca ,&nbsp;Marco Caramello ,&nbsp;Matteo Di Prinzio ,&nbsp;Francesca Giacobbo ,&nbsp;Laura Savoldi ,&nbsp;Antonio Cammi","doi":"10.1016/j.fusengdes.2025.115080","DOIUrl":"10.1016/j.fusengdes.2025.115080","url":null,"abstract":"<div><div>Magnetohydrodynamics (MHD) studies the dynamics of electrically conducting fluids under the influence of a magnetic field and it is relevant in several nuclear applications. However, the high computational cost of multi-physics MHD simulations poses a challenge. Reduced Order Modelling (ROM) offers a promising alternative, enabling lower-dimensional approximations while preserving accuracy. This allows for a reduction in the computational time and, at the same time, accurate approximations of the intricate physics involved in fusion reactors. However, ROM techniques are relatively new within the MHD framework, and benchmark test cases should be considered in this first stage for verification and validation. Therefore, this study applies the ROM methodology to a MHD scenario to study their potentialities (and eventual criticalities) for this class of problems. The benchmark test case considered in this work is the Backward-Facing Step. The obtained results contribute to assessing the capabilities of ROM methodologies in MHD scenarios, demonstrating their potential to enhance computational efficiency in this field and representing a critical step towards advancing the computational modelling of complex systems in nuclear fusion.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115080"},"PeriodicalIF":1.9,"publicationDate":"2025-04-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143873716","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Thermomechanical failure analysis of IFMIF-DONES target under off-nominal extreme conditions","authors":"M.A. Vázquez-Barroso ,&nbsp;C. Torregrosa-Martín ,&nbsp;J. Maestre","doi":"10.1016/j.fusengdes.2025.115068","DOIUrl":"10.1016/j.fusengdes.2025.115068","url":null,"abstract":"<div><div>IFMIF-DONES will be a radiological facility for material irradiation replicating conditions expected in future fusion reactors. It will employ a 40 MeV deuteron beam directed at a liquid Li target circulating at 15 m/s to generate high-energy neutrons, depositing 5 MW. The Back-Plate (BP), placed immediately downstream of the Li, separates the vacuum of the accelerator and target chambers from the low-pressure He atmosphere housing the irradiation modules. A critical scenario postulates an eventual loss of liquid Li curtain thickness without shutting down the beam, risking a direct or partial deuteron beam impact on the BP causing large power deposition. This study provides the BP dynamic thermomechanical response, aiming at characterizing the involved timings in the impact-triggered events, such as mechanical failure, melting or vaporization. This is important to evaluate the eventual mobilization of the BP volatilized activated material and the available timings for beam shutdown. The methodology involves Monte-Carlo simulations for power deposition data integrated into a Finite Element model in <span>ANSYS</span> for transient thermal and structural analyses. Results include timings for melting, vaporization, and mechanical response as function of the beam footprint area and the Li jet thickness reduction.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115068"},"PeriodicalIF":1.9,"publicationDate":"2025-04-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143873714","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Actuator management for the first ITER plasma operation campaign","authors":"O. Kudláček ,&nbsp;W. Treutterer ,&nbsp;I. Gomez Ortiz ,&nbsp;P.T. Lang ,&nbsp;R. Nouailletas ,&nbsp;L. Pangione ,&nbsp;M. Reich ,&nbsp;S. Stanek ,&nbsp;P. de Vries ,&nbsp;A. Vu ,&nbsp;D. Weldon ,&nbsp;L. Zabeo","doi":"10.1016/j.fusengdes.2025.115071","DOIUrl":"10.1016/j.fusengdes.2025.115071","url":null,"abstract":"<div><div>An important goal of the Start of Research Operation (SRO) at ITER is the commissioning of the Plasma Control System (PCS) and the testing of algorithms for heating and MHD, shape, error field, density, reaching H-mode, and ELM control. The ITER PCS will utilize magnetic actuators (error field coils, poloidal field coils), fueling actuators (4 pellet injectors connectable to 6 pellet flight tubes, 60 gas valves), and heating actuators (48 gyrotrons connectable to 9 steerable mirrors and 1 ICRH antenna). This contribution focuses on the development of the Actuator Management (AM) for the fueling and heating actuators.</div><div>The purpose of the AM is to link controllers to multi-purpose actuators in order to properly manage conflicting requests, simplify the PCS architecture and increase its robustness. It is achieved by aggregating all of the fueling and heating actuators that have a similar impact on the plasma into Virtual Actuators (VA), which are treated as single entities by the controllers. The VAs provide the actuation limits to the controllers and distribute the controller command amongst their member actuators. It simplifies the controller design, increases its actuation amplitude, and robustness against actuator faults.</div><div>This contribution details the generic AM architecture and its extensions for specific features of the above listed control tasks using the actuators that are foreseen at ITER. In particular, we will present a method to treat actuators with steerable launchers, and a way to orchestrate pellet firing for both fueling and ELM pacing control tasks.</div><div>The presented algorithms are to be implemented into the ITER PCS and used for its operation in the SRO phase.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115071"},"PeriodicalIF":1.9,"publicationDate":"2025-04-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143873713","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Fins: Improving tritium extraction systems and permeation sensors with the adoption of extended surfaces","authors":"Ciro Alberghi ,&nbsp;Luigi Candido ,&nbsp;Daniele Martelli ,&nbsp;Francesca Papa ,&nbsp;Marco Utili ,&nbsp;Alessandro Venturini","doi":"10.1016/j.fusengdes.2025.115065","DOIUrl":"10.1016/j.fusengdes.2025.115065","url":null,"abstract":"<div><div>Tritium extraction from lithium-lead (PbLi, 15.7 at. % Li) and tritium concentration measurement in the eutectic alloy represent some of the most challenging aspects of the R&amp;D activities aimed to the development of ITER and the European DEMO reactor. To efficiently design Permeator Against Vacuum (PAV) systems and Hydrogen isotopes Permeation Sensors (HPS), theoretical models for the evaluation of the permeation flux have been proposed in literature, but methodologies for the improvement of their performances are still lacking. In this paper a new concept of finned permeator is analysed, leveraging the analogy between mass transport and heat transfer. In PAV and HPS, the low-pressure side is usually kept under medium/high vacuum conditions and surface phenomena can play an important role, especially when the membrane presents oxidation. The fin approach is particularly effective in these cases, where transport kinetics is dominated by surface effects (diffusion in the bulk is relatively fast) and can be used as a method to increase the permeation of hydrogen isotopes with limited increase in system size. Within the paper, the mathematical model for the extended surface placed on the vacuum side is derived and simple relations for design parameters for the finned surface, like fin efficiency and effectiveness, are derived. The solution of this analytical model is compared with numerical results for a PAV system with niobium membrane under relevant conditions for DEMO reactor.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115065"},"PeriodicalIF":1.9,"publicationDate":"2025-04-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143865020","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Restoration in drawn tungsten wires of tungsten fiber-reinforced tungsten composites","authors":"Daniel Ahlin Heikkinen Wartacz ,&nbsp;Johann Riesch ,&nbsp;Karen Pantleon ,&nbsp;Wolfgang Pantleon","doi":"10.1016/j.fusengdes.2025.115038","DOIUrl":"10.1016/j.fusengdes.2025.115038","url":null,"abstract":"<div><div>Fusion energy holds great promise as a sustainable solution to meet global energy demands, offering a quasi-inexhaustible, secure and environmentally friendly energy resource. Materials facing the burning plasma in fusion reactors must withstand extreme conditions. Tungsten, the current choice for plasma-facing materials, is at risk of embrittlement if exposed to high temperatures due to restoration processes increasing its ductile-to-brittle transition temperature. This embrittlement limits operation of tungsten under fusion-relevant conditions. Tungsten fiber-reinforced tungsten (W<span><math><msub><mrow></mrow><mrow><mi>f</mi></mrow></msub></math></span>/W) composites are developed to mitigate brittleness and achieve pseudo-ductile behavior, utilizing drawn, potassium-doped tungsten fibers embedded in a pure tungsten matrix to enhance toughness compared to pure tungsten significantly. In view of the high heat fluxes and the expected high steady-state operation temperatures in a fusion reactor, thermal stability of the plasma-facing material becomes crucial. Model W<span><math><msub><mrow></mrow><mrow><mi>f</mi></mrow></msub></math></span>/W systems containing a single tungsten fiber in a dense tungsten matrix, with or without an yttria interlayer, are annealed at 1450 °C for up to 2 days to evaluate their thermal stability. As the tungsten fibers are primarily responsible for the pseudo-ductile behavior, this investigation focuses on analyzing the thermal stability of the fibers and their immediate vicinity in the surrounding matrix. Changes within the tungsten fibers, including alterations of boundary spacing and crystallographic texture, are analyzed using electron backscatter diffraction (EBSD) and further post-processing of the orientation data. Quantification shows a substantial increase in boundary spacing in the fibers after annealing. This indicates recrystallization, where many boundaries with disorientation angles up to 50°are removed. The crystallographic texture in the fibers changes only slightly during annealing, if at all.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115038"},"PeriodicalIF":1.9,"publicationDate":"2025-04-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143869553","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Application of the Gaede effect to the neutralizer of DTT neutral beam injector","authors":"Fabio Veronese ,&nbsp;Piero Agostinetti ,&nbsp;Andrea Murari","doi":"10.1016/j.fusengdes.2025.115077","DOIUrl":"10.1016/j.fusengdes.2025.115077","url":null,"abstract":"<div><div>Neutral Beam Injectors (NBIs) are key components in the additional heating of the tokamak plasmas. In this scheme, fast charged ions are accelerated electrostatically by suitable grids and then neutralized through a gas target in a dedicated neutralizer, in order to penetrate the strong confining magnetic field. NBIs require a powerful Gas Vacuum System (GVS) in order to operate, and any simplification to this element has a positive impact on the wall-plug efficiency of the NBI, as well as on the overall procurement. In this paper, new design concepts that rely on the Gaede effect are proposed for the neutralizer: their objective is the reduction of the gas conductance through the component by means of appropriate shaping of the internal walls, while maintaining enough clearance for the fast beam to pass through. In this way, the same gas target can be achieved with a reduced neutral gas input, decreasing the throughput that the GVS must be able to evacuate. This work gives a description of the application of this concept to the Divertor Tokamak Test (DTT)’s own NBI, which has been carried out by comparing different design options by means of several simulations with various vacuum codes. Due to the significant improvement that the Gaede effect is expected to bring to the NBI gas economy, this design application has been patented and will be proposed to be implemented in the NBI of the DTT facility, under construction in Frascati (Italy).</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115077"},"PeriodicalIF":1.9,"publicationDate":"2025-04-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143869552","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Corrosion behavior of 316L austenitic steel in liquid lithium lead alloy under stress loading conditions","authors":"Masatoshi Kondo ,&nbsp;Yota Koike ,&nbsp;Kenshi Ohno ,&nbsp;Kaoru Omiya ,&nbsp;Minho Oh ,&nbsp;Naoko Oono-Hori","doi":"10.1016/j.fusengdes.2025.115115","DOIUrl":"10.1016/j.fusengdes.2025.115115","url":null,"abstract":"<div><div>The material compatibility is one of the important issues for the development of liquid LiPb blanket of fusion reactors. The corrosion characteristics of 316L austenitic steel under tensile stress loadings were investigated by means of the immersion of C-ring specimens into liquid LiPb at 573 K and 773 K for 1000 h. The corrosion of 316L austenitic steel was negligibly small in liquid LiPb at 573 K regardless of the tensile stress loadings. The corrosion was caused by the dissolution of alloying elements (i.e., Fe, Cr and Ni) on the surface of specimens exposed to liquid LiPb at 773 K without the tensile stress loadings. The phase transformation from fcc structure of austenite to bcc structure of ferrite was induced on the surface according to the depletion of Ni in the corrosion procedure, which was austenite-stabilizing element. The dissolution corrosion was mitigated on the area around the apex of the C-ring specimens where the tensile stresses of 100 MPa and 200 MPa were applied. Cr- and C- rich regions were detected on the tensile-strained area of the specimens which could function as an anti-corrosion layer in liquid LiPb. The enrichment of Cr was induced by the depletion of Fe and Ni on the surface since the solubility of Cr was smaller than that of Fe and Ni in liquid LiPb. The dissolution of Cr could be mitigated by the carburization on the surface. The stress-induced martensitic transformation and the introduction of dislocations could contribute the formation of Cr- and C- rich regions.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115115"},"PeriodicalIF":1.9,"publicationDate":"2025-04-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143864974","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Design of the cryogenic distribution system for the ITER disruption mitigation based on shattered pellet injection","authors":"S. Giors ,&nbsp;F. Adong ,&nbsp;O. Barana ,&nbsp;A. Barturen Montes ,&nbsp;F. Dhalla ,&nbsp;S. Jachmich ,&nbsp;V. Kulaev ,&nbsp;U. Kruezi ,&nbsp;N. Luchier ,&nbsp;J. Manzagol ,&nbsp;F. Millet ,&nbsp;M. Parekh ,&nbsp;A. Rizzato","doi":"10.1016/j.fusengdes.2025.115116","DOIUrl":"10.1016/j.fusengdes.2025.115116","url":null,"abstract":"<div><div>The ITER tokamak will be equipped with a machine protection system to avoid or mitigate the damage to its in-vessel components in the event of plasma disruptions. The Disruption Mitigation System (DMS) will be based on Injection of Shattered Pellets (SPI) made of hydrogen, neon or their mixtures into the plasma, to convert the plasma energy into radiation while avoiding the formation or dissipate the energy of runaway electrons and to minimize the electromagnetic loads by controlling the plasma current quench.</div><div>To achieve the disruption mitigation requirements and fulfill the pulse rate for the ITER Research Plan, the DMS Cryogenic Distribution System (CDS) shall form cylindrical pellets with a diameter of 28.5 mm, a length of 57 mm and good integrity, by de-sublimation of gases inside a Supercritical helium (SHe) cooled Cold Cell (CC), in ≈1200 s (20 min) for hydrogen, and maintain their availability over several plasma pulses.</div><div>The DMS CDS was integrated into the ITER baseline at a late design stage, with limited SHe cooling capacity supplied in parallel to the cryopumps for vacuum vessel, cryostat and neutral beam injectors. Seven Cold Distribution Boxes (CDBs) dedicated to the DMS equatorial and upper port locations were introduced, each equipped with a Joule-Thompson (JT) expansion valve and a liquid helium vessel, to supply the SHe flow to 27 CCs at a stable temperature of ∼5 K for pellet formation and preservation. The CC design was supported by de-sublimation numerical modelling and experiments to optimize the pellet shape and integrity and to minimize the CC cooling requirement to form pellets within an acceptable time. The cryogenic system design aimed at minimizing heat losses while considering the very challenging environmental (magnetic field, nuclear, seismic) and complex integration requirements.</div><div>This paper presents the DMS CDS description, following the final design review in 2024, focusing on the CC novel design supported by CFD models and laboratory experiments.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115116"},"PeriodicalIF":1.9,"publicationDate":"2025-04-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143869554","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Achievement of first plasma and successful integrated commissioning in JT-60SA","authors":"K. Takahashi ,&nbsp;JT-60SA Integrated Project Team","doi":"10.1016/j.fusengdes.2025.115059","DOIUrl":"10.1016/j.fusengdes.2025.115059","url":null,"abstract":"<div><div>Integrated Commissioning of JT-60SA, which will be the world’s largest tokamak until ITER operates, was completed in 2023. The device is designed to operate under good vacuum condition, which is by nature high insulation, and all superconducting magnets could withstand the operation voltage even if they were not Paschen proof. Applying the fast interlock system detecting vacuum deterioration inside the cryostat and the consideration of limited operation voltage (&lt;2kV) of PF magnets, the integrated commissioning of JT-60SA was restarted in May 2023. The pump-down of the cryostat and vacuum vessel and the cool-down of the magnets were subsequently carried out. Taking advantage of the experience of the first cool-down, its speed was enhanced by increasing helium flow. In addition, the vacuum vessel was baked at 200 °C for wall cleaning during cool-down to save time. Furthermore, glow discharge operation was also executed as another process of wall clearing. After the high voltage (HV) test of all magnets at the operation condition, which were high vacuum (∼10^-6 Pa) and cryogenic temperature (∼4K), the magnet energization was successfully carried out and followed by plasma operation. The first plasma with limiter configuration and plasma current of 130kA was obtained applying Trapped Particle Configuration (TPC) on Oct. 23, 2023. Electron Cyclotron Heating (ECH) assisted plasma breakdown was performed with a low toroidal electric field (∼0.15 V/m).</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115059"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143865019","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Thermo-structural analysis on the DTT ECH Transmission Line Single-Beam mirrors","authors":"A. Salvitti ,&nbsp;A. Moro ,&nbsp;A. Bruschi ,&nbsp;G. Calabrò ,&nbsp;F. Fanale ,&nbsp;P. Fanelli ,&nbsp;S. Garavaglia ,&nbsp;G. Granucci ,&nbsp;S. Meloni ,&nbsp;P. Platania ,&nbsp;A. Romano","doi":"10.1016/j.fusengdes.2025.115088","DOIUrl":"10.1016/j.fusengdes.2025.115088","url":null,"abstract":"<div><div>The Electron Cyclotron Heating (ECH) of the Divertor Tokamak Test (DTT) facility is its main system to heat the plasma with an installed power of 32 MW. A cluster of the ECH consists of 1 MW/170 GHz sources (8 gyrotrons), a quasi-optical Transmission Line (TL) and two antennas. The TL efficiently transmits the microwave beams, produced by gyrotrons, up to the launchers. Based on the quasi-optical concept, the TL employs mirrors for the transmission of the beams. One of the three segments in which it is conceptually divided is a Single-Beam section in the Gyrotron hall (SBG), where the beams propagate with an independent optical path. A small fraction of beam power is absorbed by the reflecting mirror due to ohmic losses, resulting in a total power absorbed by an SBG mirror of ∼1-2 kW. The mirror temperature increases, generating deformations. The loss of the nominal shape of the mirror can contribute to reducing the transmission efficiency, compromising the nominal parameters that ensure the effective performance of the beam propagation system. The conceptual design of different SBG mirrors is carried out, simulating the thermal-structural behaviour. The mirrors, actively water-cooled through channels designed as elongated spirals with variable pitch, achieve maximum temperatures of 28 °C and low deformations (∼23 μm), demonstrating the effectiveness of the cooling system. In the end, the same method is then applied in a parametric analysis of the thickness to evaluate the best performance, if necessary. It showed that the initial assumption on the thickness appears to be a well-founded choice.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115088"},"PeriodicalIF":1.9,"publicationDate":"2025-04-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143860118","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}