Fusion Engineering and Design最新文献

英文中文

Enhanced deuterium extraction efficiency from lithium-lead droplets in a vacuum

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-03 DOI: 10.1016/j.fusengdes.2025.114917

Fumito Okino , Yukinori Hamaji , Juro Yagi , Teruya Tanaka

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.

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引用次数: 0

First thermo-structural vacuum barrier design for the EU DEMO feeders 为欧盟 DEMO 送料机设计的首个热结构真空屏障

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-01 DOI: 10.1016/j.fusengdes.2025.114905

Corrado Groth, Andrea Chiappa, Marco Evangelos Biancolini

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.

引用次数: 0

Tritium permeation through Inconel 600 under high temperature, high pressure water environment: Influence of oxidation of coexisting materials and gas addition

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-01 DOI: 10.1016/j.fusengdes.2025.114896

Azusa Matsumoto , Yuji Hatano

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.

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引用次数: 0

Mirror dual cleaning of ITER equatorial diagnostic Wide Angle Viewing System

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-03-01 DOI: 10.1016/j.fusengdes.2025.114892

L. Marot , P. Hiret , S. Dine , A. Dmitriev , L. Letellier , S. Vives , F. Le Guern , J. Piqueras , M. Martina , R. Steiner , R. Maffiolini , A. Tonin , E. Meyer

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.

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引用次数: 0

Parametric study of neutral beam injection heating and current drive in NCST

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-02-28 DOI: 10.1016/j.fusengdes.2025.114897

X.X. Zhang, X.C. Chen, S.Q. Liu, H. Chen, X.F. Wu

In order to study the physics associated with high-

β

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.

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引用次数: 0

Supply and demand of tungsten in a fleet of fusion power plants 核聚变发电厂的钨供需情况

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-02-28 DOI: 10.1016/j.fusengdes.2025.114881

E. Day-San , G.C. Blackett , M. Dornhofer , A.K. Manduku , M.D. Anderton , L. Tanure , T.P. Davis

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 (W₂B), 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.

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引用次数: 0

Advances in magnet and shielding designs for fusion and high energy physics applications

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-02-27 DOI: 10.1016/j.fusengdes.2025.114899

L. Giannini , C. Luongo , L. Bottura , B. Bordini , A. Lechner , A. Kolehmainen , D. Leichtle , A. Portone , P. Testoni , J. Bajari , M. Siccinio , C. Bachmann , G. Federici

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.

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).

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.

{"title":"Advances in magnet and shielding designs for fusion and high energy physics applications","authors":"L. Giannini , C. Luongo , L. Bottura , B. Bordini , A. Lechner , A. Kolehmainen , D. Leichtle , A. Portone , P. Testoni , J. Bajari , M. Siccinio , C. Bachmann , 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}

引用次数: 0

Distributed editing of experiment programs at Wendelstein 7-X

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-02-27 DOI: 10.1016/j.fusengdes.2025.114880

Anett Spring, Heike Riemann, Marc Lewerentz, W7-X Team

Configuring experiment programs at complex fusion experimental facilities such as Wendelstein 7-X is a challenge: the amount of parameters, the growing number of components, the required specialized knowledge of technical systems and diagnostics, the complex timing – all this demands an integrated solution for experiment leaders, engineers, and physicists.With the growing number of integrated components, the W7-X experiment program editor Xedit was extended to cope with TaskLinks: allowing externally configured program parts of individual components to be linked into the planned sequence of the central experiment program. The preparation of the specific Tasks is done by the component owners using a local Xedit instance – in the same way, as the users are already familiar with from the creation of local programs for commissioning or calibration runs. Centrally, as with all integrated components, the linked components’ Tasks are then visualized and all parameters are checked for limit violations or other pre-defined rules before saving the complete planned program ready for execution.

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引用次数: 0

Dose assessment for radioactive products distributed on liquid Li loops of fusion neutron sources

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-02-27 DOI: 10.1016/j.fusengdes.2025.114901

Shunsuke Kenjo, Makoto Oyaidzu, Saerom Kwon, Kentaro Ochiai, Satoshi Sato

Radiation dose rates around liquid Li loops in International Fusion Materials Irradiation Facility (IFMIF)-like fusion neutron sources, such as A-FNS, increase because of the radionuclides produced by high-energy deuterons and neutrons. This study assesses radiation dose rates around the pipe, heat exchanger (HX), and electromagnetic pump (EMP) of a liquid Li loop using the Monte Carlo code MCNP6.2 and the photon data library mcplib84 to advance the development of maintenance strategies. The radiation sources include not only major nuclides (tritium and Be-7) reported in previous studies but also other radionuclides that have not been well investigated thus far. The calculation results suggest that hands-on maintenance of Li pipes can be performed after the Li drain. However, accesses HX and EMP are limited because of nuclides such as Y-88, which are produced from impurities in liquid Li, even when Be-7 is trapped in Cold Trap.

引用次数: 0

MOOSE-based Tritium Migration Analysis Program, Version 8 (TMAP8) for advanced open-source tritium transport and fuel cycle modeling

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2025-02-27 DOI: 10.1016/j.fusengdes.2025.114874

Pierre-Clément A. Simon , Casey T. Icenhour , Gyanender Singh , Alexander D. Lindsay , Chaitanya Bhave , Lin Yang , Adriaan Riet , Yifeng Che , Paul Humrickhouse , Pattrick Calderoni , Masashi Shimada

Tritium management is critical for the safety, sustainability, and economics of fusion energy systems, and advanced and reliable modeling tools help accelerate the development of tritium technologies. This paper presents the Tritium Migration Analysis Program, Version 8 (TMAP8), an open-source, MOOSE-based application developed to provide state-of-the-art tritium transport and fuel cycle modeling capabilities. TMAP8 aims to expand the capabilities of previous versions (i.e., TMAP4 and TMAP7) by leveraging modern computational techniques, ensuring high software quality assurance standards (key to building trust), and enabling multispecies, multiscale, and multiphysics simulations for integrated tritium transport modeling in complex geometries. This paper outlines TMAP8’s scope and rigorous development practices, emphasizing its transparency, accessibility, modularity, and reliability. We present the current suite of verification and validation cases based on those from TMAP4, demonstrating TMAP8’s accuracy and reliability against analytical solutions and experimental data. Additionally, the paper showcases TMAP8’s integrated fuel cycle modeling capabilities, highlighting its applicability at various scales and levels. The TMAP8 code and documentation are openly available, promoting collaborative development and widespread adoption within the fusion community. Future work will soon expand TMAP8’s verification and validation suite to include those from TMAP7 and other recent experimental studies for validation.

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