Fusion Engineering and Design最新文献

{"title":"Design of a 170GHz quasi-optical notch filter for microwave-based diagnostics protection on the burning plasma device","authors":"Chen Luo ,&nbsp;Xiaopin Tang ,&nbsp;Calvin Domier ,&nbsp;Xiaoliang Li ,&nbsp;Ang Ti ,&nbsp;Pengjun Sun ,&nbsp;Shasha Qiu ,&nbsp;Xinhang Xu ,&nbsp;Xianzi Liu ,&nbsp;Yilun Zhu ,&nbsp;Neville Luhmann","doi":"10.1016/j.fusengdes.2025.114925","DOIUrl":"10.1016/j.fusengdes.2025.114925","url":null,"abstract":"<div><div>170 GHz gyrotrons will be used for Electron Cyclotron Heating (ECH) on the under-development burning plasma device, with a maximum power output of 16 MW with center frequency of 170 GHz. The reliable 170 GHz quasi-optical notch filters protect microwave-based diagnostic instruments from unabsorbed ECH power. This design utilizes a metal circular periodic structure, which is straightforward to manufacture, to achieve reflection and absorption of 170 GHz electromagnetic waves transmitted through space, thereby achieving the notch filtering effect (attenuation &gt; 60 dB). Unlike the rectangular ring design used in low-frequency quasi-optical notch filters, the circular design caters to the shorter wavelength of 170 GHz, which necessitates reduced fabrication error tolerance. To achieve both high quality filter and higher processing yields, we compensate for the performance degradation of the notch filter caused by processing errors by optimizing the controllable metal circle radius and incident angle. This paper discusses the impact of material size selection and machining accuracy deviations on the performance of the notch filter.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"214 ","pages":"Article 114925"},"PeriodicalIF":1.9,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143552669","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":"Thermal-hydraulic scaling of the prototypical mock-up for European DEMO HCPB breeding blanket","authors":"Gaurav Verma,&nbsp;Guangming Zhou,&nbsp;Francisco A. Hernández","doi":"10.1016/j.fusengdes.2025.114924","DOIUrl":"10.1016/j.fusengdes.2025.114924","url":null,"abstract":"<div><div>The Helium-Cooled Pebble Bed (HCPB) blanket concept is a leading candidate for the driver blanket in the European Fusion Demonstration Power Plant (DEMO), developed within the EUROfusion framework's Work Package Breeding Blanket (WPBB). A crucial component of the HCPB blanket is the First Wall (FW), which must withstand high heat fluxes from the plasma while maintaining a uniform temperature distribution as much as possible. This uniformity is achieved through an alternating coolant flow design. The present study focuses on the thermal-hydraulic scaling methodology for the First Wall-Prototype Mock-Up (FW-PMU) to replicate the HCPB blanket's thermal and hydraulic performance. The FW-PMU is designed to emulate the flow distribution and thermal behavior of the HCPB Breeding Blanket (BB) First Wall, including flow patterns from the Breeder Zone Manifold to the fuel-breeder pins. In the present work, a power-to-volume scaling approach is employed to maintain the characteristic time ratio by analyzing non-dimensional parameters, while scaling of component dimensions in the breeder zone manifold ensures a consistent velocity field. The FW-PMU is set for testing in HELOKA, a high-pressure, high-temperature helium facility at the Karlsruhe Institute of Technology to validate the HCPB First Wall design under realistic operating conditions, advancing its potential deployment in fusion energy systems.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"214 ","pages":"Article 114924"},"PeriodicalIF":1.9,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143533545","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Conceptual design of inductive magnetic sensors using photolithography processes","authors":"Duccio Testa ,&nbsp;Marcus Cemes ,&nbsp;Eugénie Decaux ,&nbsp;Camille Lavilla ,&nbsp;Florian Tanguy ,&nbsp;Mehdi Amor","doi":"10.1016/j.fusengdes.2025.114907","DOIUrl":"10.1016/j.fusengdes.2025.114907","url":null,"abstract":"<div><div>Inductive magnetic sensors are needed for the operation of fusion devices to monitor high frequency (HF) fluctuations and as a back-up to the low-frequency (LF, still inductive) magnetic sensors used for the measurements leading to the reconstruction of the equilibrium. The technical specifications for these two types of inductive magnetic sensors are rather different: this is a major conceptual difficulty in the design of inductive magnetic sensors, and then most often different sets of inductive magnetic sensors are used, which significantly complicates R&amp;D activities, prototyping and manufacturing.</div><div>Most (∼500) of the inductive magnetic sensors currently being deployed in ITER have been produced with the Low-Temperature CoFired Ceramic (LTCC) technology. The LTCC technology is now at least 20 years old and new processes have been developed for industrial applications, essentially based on different photolithography (PL) processes. The main advantage of the PL processes is that a much smaller track width can be achieved: a smaller dd1 allows to “pack” more planar winding loops (=m) enclosing a larger area over a smaller geometrical surface, thus keeping the same overall effective are NA_EFF∝m while significantly reducing the sensors’ self-inductance L_SELF∝m². Then, with PL techniques the same design could in principle be used for both high-frequency and low-frequency applications, the difference simply being the number of stacked-up layers (=<em>n</em>) used to make-up the entire sensor.</div><div>Continuing from our earlier work, in this paper we will present recent advances in our processes for producing inductive magnetic sensors using PL methods, most notably on the use of synthetic Sapphire wafers, on increasing the track thickness, on developing multi-layers sensors, on producing miniaturized saddle loops, and finally on developing packaging solutions for installing these inductive magnetic sensors in-vessel and ex-vessel.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"214 ","pages":"Article 114907"},"PeriodicalIF":1.9,"publicationDate":"2025-03-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143533544","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Enhanced deuterium extraction efficiency from lithium-lead droplets in a vacuum","authors":"Fumito Okino ,&nbsp;Yukinori Hamaji ,&nbsp;Juro Yagi ,&nbsp;Teruya Tanaka","doi":"10.1016/j.fusengdes.2025.114917","DOIUrl":"10.1016/j.fusengdes.2025.114917","url":null,"abstract":"<div><div>The authors conducted tritium extraction from falling liquid lithium-lead (LiPb) droplets in a fusion blanket loop. We reported an extraction efficiency above 0.6 from LiPb droplets of diameter 1.89 mm at 15th International Symposium on Fusion Nuclear Technologies (ISFNT15_2023). Even though it is still below the European Demonstration Power Plant (EU-DEMO) design criteria of above 0.8 which is commonly recognized as requisite minimum. Furthermore, many droplet nozzles are required to attain LiPb flow rate design criteria. Therefore, the aim is to increase the efficiency and flow rate. Enlarging the droplet diameter is a simple way to boost the flow rate. However, this degrades the efficiency. To address this drawback, we considered the tandem extraction method, which consists of two extraction processes. In the first process, dissolved deuterium is extracted from the partially filled LiPb flow in an inlet pipe under a vacuum. The second process entails the conventional extraction from falling droplets. Experimental verifications and theoretical analyses were performed at the liquid metal experimental test loop installed at the National Institute for Fusion Science, at Toki Japan. A droplet size of 2.27 mm, 1.44 times the flow rate than the previous 1.89 mm, is applied to verify the process. The obtained overall efficiencies were between 0.75and 0.95, exceeding the estimated range of 0.65–0.85. The matured turbulent flow at the nozzle boosted the release of droplets, along with the tandem extraction. Other enhancement factors such as droplet break-up and surface oscillation were considered as scarce effects. Further verifications are inevitable even when the results suggest a high-efficiency extraction feasibility from larger droplets.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"214 ","pages":"Article 114917"},"PeriodicalIF":1.9,"publicationDate":"2025-03-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143533543","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"First thermo-structural vacuum barrier design for the EU DEMO feeders","authors":"Corrado Groth,&nbsp;Andrea Chiappa,&nbsp;Marco Evangelos Biancolini","doi":"10.1016/j.fusengdes.2025.114905","DOIUrl":"10.1016/j.fusengdes.2025.114905","url":null,"abstract":"<div><div>The vacuum barrier (VB) is designed to separate the feeder into two distinct vacuum regions: the main cryostat vacuum and the feeder vacuum. This separation enhances thermal insulation and facilitates maintenance and access to feeder components. Beyond sustaining pressure under both normal operation and potential malfunctions, the VB also minimizes heat transfer from the environment to low-temperature systems. This paper details the optimization process of an initial VB design, utilizing a Radial Basis Functions-based mesh morphing approach. Shape variations were applied concurrently to structural and thermal simulations, enabling parameterization of the complex, coupled nonlinear system where the structural model withstands both pressure and temperature loads. The optimal VB configuration, meeting structural and thermal criteria, was ultimately identified through response surface optimisation.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"214 ","pages":"Article 114905"},"PeriodicalIF":1.9,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143519220","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":"Tritium permeation through Inconel 600 under high temperature, high pressure water environment: Influence of oxidation of coexisting materials and gas addition","authors":"Azusa Matsumoto ,&nbsp;Yuji Hatano","doi":"10.1016/j.fusengdes.2025.114896","DOIUrl":"10.1016/j.fusengdes.2025.114896","url":null,"abstract":"<div><div>Tritium (T) permeation through steam generator piping from the primary to the secondary side of a water-cooled breeding blanket system increases a risk of exposure of workers and members of the public. From this viewpoint, the T permeation through Inconel 600, a candidate material of steam generator piping, under exposure to tritiated water was examined at 280 °C and 6.4 MPa by focusing attention on the influence of oxidation of coexisting materials and that of H₂ and O₂ gas addition. The T permeation rate through Inconel 600 sample was sensitively dependent on the oxidation rate of coexisting material, and a high permeation rate was observed with a material with high oxidation rate. The H₂ gas addition also resulted in a remarkable increase in T permeation rate, while the O₂ gas addition led to clear reduction. These observations indicated that HT generated by the oxidation of coexisting material by HTO and the isotope exchange reaction between HTO and H₂ gas (HTO + H₂ → H₂O + HT) contributed to the permeation. Reduction in T permeation in a steam generator appears possible by minimizing oxidation of coexisting materials in the primary loop and/or continuous O₂ gas supply.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"214 ","pages":"Article 114896"},"PeriodicalIF":1.9,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143527007","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Mirror dual cleaning of ITER equatorial diagnostic Wide Angle Viewing System","authors":"L. Marot ,&nbsp;P. Hiret ,&nbsp;S. Dine ,&nbsp;A. Dmitriev ,&nbsp;L. Letellier ,&nbsp;S. Vives ,&nbsp;F. Le Guern ,&nbsp;J. Piqueras ,&nbsp;M. Martina ,&nbsp;R. Steiner ,&nbsp;R. Maffiolini ,&nbsp;A. Tonin ,&nbsp;E. Meyer","doi":"10.1016/j.fusengdes.2025.114892","DOIUrl":"10.1016/j.fusengdes.2025.114892","url":null,"abstract":"<div><div>The metallic first mirrors (FMs) are expected to play a crucial role in most optical diagnostic systems in the International Thermonuclear Experimental Reactor (ITER). However, these mirrors will be subjected to deposition of the first-wall materials, which will comprise their optical properties. Thus, the FMs will require periodic cleaning to restore their optical properties, which is anticipated to be achieved using an in situ plasma cleaning technique employing radio-frequency (RF) discharges. The left tangential line of sight of the Wide Angle Viewing System (WAVS) for the ITER Equatorial Port 12 FM unit, designed by CEA Cadarache, was simplified and manufactured at the University of Basel for RF-cleaning tests in a realistic geometry. Plasma ignition on mirrors M1 and M2 was achieved at a frequency of 13.56 MHz. Both mirrors were successfully cleaned with a 20-nm aluminum oxide (Al₂O₃) film replicating a contaminant layer. However, powering M1 and M2 in a dual cleaning regime with argon atmosphere at 1 Pa with 100 W of RF power for 1h and 45min (with a −300 V self-bias) allowed for the simultaneous removal of the Al₂O₃ layer from the rhodium mirror insets on both mirrors. The temperature of the mineral insulating (MI) cables used for powering the first mirror remained well below the maximum rated temperature. A cleaning demonstration was also performed on a stratified mirror prototype developed by CEA. The effectiveness of the dual cleaning technique for the Equatorial Port 12 WAVS diagnostic was validated by powering both FMs in a direct-current (DC) coupled regime, as all reported diagnostics will use a DC decoupled mode.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"214 ","pages":"Article 114892"},"PeriodicalIF":1.9,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143519355","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Supply and demand of tungsten in a fleet of fusion power plants","authors":"E. Day-San ,&nbsp;G.C. Blackett ,&nbsp;M. Dornhofer ,&nbsp;A.K. Manduku ,&nbsp;M.D. Anderton ,&nbsp;L. Tanure ,&nbsp;T.P. Davis","doi":"10.1016/j.fusengdes.2025.114881","DOIUrl":"10.1016/j.fusengdes.2025.114881","url":null,"abstract":"<div><div>To enable the widespread adoption of nuclear fusion power plants, a reliable tungsten supply chain is essential for plasma-facing and radiation shielding components in spherical and D-shaped tokamaks. The ARIES-ST and EU-DEMO1 design points were used as the basis for neutronic modelling to evaluate tungsten consumption during 40 full-power years (fpy) at 500MWth and 2,000MWth fusion powers. Four materials were considered for radiation shielding: ITER Grade W, tungsten carbide (WC), tungsten boride (W<strong>₂</strong>B), and WC/Co. In spherical tokamaks, the central column radiation shielding, due to its proximity to the plasma, was found to be the primary consumer of tungsten. In contrast, the EU-DEMO1 design demonstrated minimal consumption by the shield due to increased reactor volume and shielding via the breeder blanket. Over 40 fpy, the ARIES-ST reactor consumed 4,231 tonnes of tungsten at 500MWth and 29,034 tonnes at 2,000MWth, while EU-DEMO1 consumed 3,945 tonnes at 500MWth and 9,554 tonnes at 2,000MWth, with the 2,000MWth EU-DEMO1 model being the most material efficient design in the context of a reactor roll out model. Three tungsten supply scenarios were explored, highlighting the need for new mining resources by the mid-2040s to ensure a sustainable supply for fusion plants by 2100. If the UK or US were to operate fusion power fleets without domestic tungsten sources, their supply would likely fall drastically short without heavy investment and expansion.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"214 ","pages":"Article 114881"},"PeriodicalIF":1.9,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143519354","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":"Parametric study of neutral beam injection heating and current drive in NCST","authors":"X.X. Zhang,&nbsp;X.C. Chen,&nbsp;S.Q. Liu,&nbsp;H. Chen,&nbsp;X.F. Wu","doi":"10.1016/j.fusengdes.2025.114897","DOIUrl":"10.1016/j.fusengdes.2025.114897","url":null,"abstract":"<div><div>In order to study the physics associated with high-<span><math><mi>β</mi></math></span> plasma and fast ions, we are exploring auxiliary heating method suitable for NanChang Spherical Tokamak(NCST). The NUBEAM code is used for numerically simulate NCST neutral beam injection (NBI) to provide relevant physical predictions. The NBI injection geometry is optimized to maximize deposition and current drive, thus avoiding excessive losses. The effects of density, electron temperature and plasma current on heating and current drive are discussed. The results show that the deposited power increases and the loss decreases with the increase of plasma density and current. The electron temperature primarily influences the fraction of power deposited on ion and electron, and the increase of electron temperature leads to a higher proportion of power deposited on ions. Lower electron density and higher electron temperature are favorable for neutral beam current drive. Under the same power conditions, the efficiency of low energy beam is superior to that of high energy beam, with the beam energy ranging from 15–25 keV.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"214 ","pages":"Article 114897"},"PeriodicalIF":1.9,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143519353","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":"Advances in magnet and shielding designs for fusion and high energy physics applications","authors":"L. Giannini ,&nbsp;C. Luongo ,&nbsp;L. Bottura ,&nbsp;B. Bordini ,&nbsp;A. Lechner ,&nbsp;A. Kolehmainen ,&nbsp;D. Leichtle ,&nbsp;A. Portone ,&nbsp;P. Testoni ,&nbsp;J. Bajari ,&nbsp;M. Siccinio ,&nbsp;C. Bachmann ,&nbsp;G. Federici","doi":"10.1016/j.fusengdes.2025.114899","DOIUrl":"10.1016/j.fusengdes.2025.114899","url":null,"abstract":"<div><div>This paper introduces the work scope of a collaboration among EUROfusion, CERN, and F4E to create models and design tools applicable both for fusion devices and muon accelerator magnets. The study described here is motivated by the commonality in challenges to design magnets and radiation shields applicable to compact fusion machines to serve as neutron source, and future high-energy physics experiments such as the muon collider or high intensity neutrino factories.</div><div>Both applications present design challenges revolving around the trade-off between the need for very high fields over a relatively large free bore, while keeping the coils sufficiently far away and properly shielded from radiation to limit heat load and damage to materials. The present focus is on creating design models and tools for superconducting solenoids built exclusively with high-temperature superconductors (HTS), or in a hybrid configuration with low-temperature superconductors (LTS).</div><div>We use the pinch solenoid magnets for a mirror fusion machine as a case study. The magnetic field requirements and configuration, material limits, structural constraints, and shielding properties are considered while using an optimization model to scan and systematically size the coil/shield combination. The investigation involves conductor selection, mechanical analyses, and cooling schemes. Moreover, we examine the optimization of magnet stability in operation, surveying radiation-shielding materials and innovative shielding concepts. The similarities in design challenges between fusion devices and accelerator magnets are highlighted.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"214 ","pages":"Article 114899"},"PeriodicalIF":1.9,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143511204","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}