Journal of Fusion Energy最新文献_第9页

Microscale Turbulence Measurements Using Doppler Reflectometer and millimeter-wave Scattering System 用多普勒反射仪和毫米波散射系统测量微尺度湍流

IF 2.1 4区工程技术

Journal of Fusion Energy Pub Date : 2025-11-10 DOI: 10.1007/s10894-025-00523-1

Tokihiko Tokuzawa, Tatsuhiro Nasu, Daiki Nishimura, Shigeru Inagaki, Akira Ejiri, Katsumi Ida, Mikirou Yoshinuma, Tatsuya Kobayashi, Kenji Tanaka, Akihide Fujisawa, Ichihiro Yamada

{"title":"Microscale Turbulence Measurements Using Doppler Reflectometer and millimeter-wave Scattering System","authors":"Tokihiko Tokuzawa, Tatsuhiro Nasu, Daiki Nishimura, Shigeru Inagaki, Akira Ejiri, Katsumi Ida, Mikirou Yoshinuma, Tatsuya Kobayashi, Kenji Tanaka, Akihide Fujisawa, Ichihiro Yamada","doi":"10.1007/s10894-025-00523-1","DOIUrl":"10.1007/s10894-025-00523-1","url":null,"abstract":"<div><p>To study microscale turbulence, two non-invasive scattering instruments that use electromagnetic waves in the microwave to millimeter-wave range have been installed at the Large Helical Device. One instrument is a Doppler reflectometer, which is suitable for observing turbulence with relatively low wavenumbers. Three circuit systems were constructed for this instrument. The Doppler reflectometer allows a very large number of spatial points (more than 30) to be observed simultaneously in the radial direction and toroidal correlation analysis to be conducted. The other instrument is a two-frequency millimeter-wave scattering system, which was developed to observe turbulence at relatively high wavenumbers. This scattering system has multiple antennas in a vacuum vessel. It can be used to study turbulence anisotropy or, in combination with the Doppler reflectometer, the response of turbulence at various scales.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"44 2","pages":""},"PeriodicalIF":2.1,"publicationDate":"2025-11-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://link.springer.com/content/pdf/10.1007/s10894-025-00523-1.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145510696","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Structural Stability of CFETR Central Solenoid Model Coil Under Assembly-Induced Offsets: Electromagnetic Loads and Instability Evaluation CFETR中央电磁模型线圈在装配诱导偏移下的结构稳定性：电磁载荷和不稳定性评估

IF 2.1 4区工程技术

Journal of Fusion Energy Pub Date : 2025-11-10 DOI: 10.1007/s10894-025-00525-z

Xianewei Wang, Wenlong Xu, Chenyang Li, Aihua Xu, Wentao Xie, Xiulian Li

{"title":"Structural Stability of CFETR Central Solenoid Model Coil Under Assembly-Induced Offsets: Electromagnetic Loads and Instability Evaluation","authors":"Xianewei Wang, Wenlong Xu, Chenyang Li, Aihua Xu, Wentao Xie, Xiulian Li","doi":"10.1007/s10894-025-00525-z","DOIUrl":"10.1007/s10894-025-00525-z","url":null,"abstract":"<div><p>Against the backdrop of the accelerated advancement of the International Thermonuclear Experimental Reactor project, the Central Solenoid Model Coil (CSMC) of the China Fusion Engineering Test Reactor (CFETR) plays a pivotal role in achieving efficient plasma confinement and control. During the assembly of CSMC, inevitable assembly errors can disrupt both the central symmetry around the central axis and the planar symmetry with respect to the mid-plane of the Nb₃Sn and NbTi coils. This symmetry disruption leads to the generation of substantial asymmetric radial and axial offset electromagnetic forces within the Nb₃Sn and NbTi coils, which potentially jeopardize the preload stability and structural integrity of supporting components. To systematically investigate the stability of CSMC under the influence of assembly errors, this study first established a calculation model based on electromagnetic field theory for three typical assembly error scenarios, followed by the computation of corresponding offset electromagnetic forces. Second, the linear instability analysis method was utilized to determine the instability modes of CSMC under different offset electromagnetic force conditions. Finally, an extreme offset state was selected for nonlinear instability analysis to quantify the critical load triggering instability. The research results indicate that, under extreme assembly errors, preload rods emerge as the weakest structural components susceptible to instability. Notably, the critical instability load demonstrates an 8.3-fold margin over the combined operational loads. This study offers critical data references for optimizing the assembly precision and conducting safety margin evaluations in CSMC operation.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"44 2","pages":""},"PeriodicalIF":2.1,"publicationDate":"2025-11-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145510695","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"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

Observation of a Fast-Ion Profile Stiffness Due To the Alfvén Eigenmode 由alfv<s:1>本征模引起的快速离子剖面刚度的观察

IF 2.1 4区工程技术

Journal of Fusion Energy Pub Date : 2025-11-08 DOI: 10.1007/s10894-025-00521-3

Shuji Kamio, Yutaka Fujiwara, Kenichi Nagaoka, Hideo Nuga, Hiroyuki Yamaguchi, Ryosuke Seki, Kunihiro Ogawa, Yasuko Kawamoto, Mitsutaka Isobe, Scott Karbashewski, Erik Granstedt, Masaki Osakabe

{"title":"Observation of a Fast-Ion Profile Stiffness Due To the Alfvén Eigenmode","authors":"Shuji Kamio, Yutaka Fujiwara, Kenichi Nagaoka, Hideo Nuga, Hiroyuki Yamaguchi, Ryosuke Seki, Kunihiro Ogawa, Yasuko Kawamoto, Mitsutaka Isobe, Scott Karbashewski, Erik Granstedt, Masaki Osakabe","doi":"10.1007/s10894-025-00521-3","DOIUrl":"10.1007/s10894-025-00521-3","url":null,"abstract":"<div><p>Fast-ion transport driven by Alfvén eigenmodes (AEs) is one critical issue facing fast-ion confinement in magnetic fusion device. In the DIII-D tokamak experiment, stiff transport of fast-ions increased with increasing neutral beam (NB) injection power when the amplitudes of multiple interacting AEs exceeded a certain threshold. These experiment results are supported by simulation studies that predict monotonically degrading fast-ion confinement and profile stiffness with increasing beam power. To investigate the universality of the fast-ion profile stiffness dependence on AE amplitude, an experiment was performed at the Large Helical Device (LHD) to scan the injection current of the NB and vary the AE amplitude. Under the experimental conditions, the AE amplitude increased linearly with NB injection power. The red shifted FIDA intensity between 663 and 665 nm, corresponding to the energy range of 98–166 keV in the ctr-direction, was used for estimating the radial profile of the fast-ion density. Evidence suggests stiffening of the fast-ion profile and degraded confinement, corroborated by a reduced neutron emission rate compared to simulations. This is consistent with the experimentally observed reduction in the expected neutron emission rate. We have demonstrated that under AE-prone confinement conditions, even if the fast-ion source increases due to NB injection, they experience enhanced transport by AEs and do not increase in density.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"44 2","pages":""},"PeriodicalIF":2.1,"publicationDate":"2025-11-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://link.springer.com/content/pdf/10.1007/s10894-025-00521-3.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145510464","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Сombined Glow + Microwave Discharges in the TOMAS Plasma Facility Сombined TOMAS等离子体设施中的辉光+微波放电

IF 2.1 4区工程技术

Journal of Fusion Energy Pub Date : 2025-11-05 DOI: 10.1007/s10894-025-00524-0

Y. P. Martseniuk, Y. V. Kovtun, A. Goriaev, D. Nicolai, K. Crombé, L. D. López-Rodríguez, J. Buermans, P. Petersson, L. Dittrich, S. Möller, S. Brezinsek

引用次数: 0

Study of Resistive Wall Mode Feedback Control with Varying Plasma Configurations 变等离子体结构的电阻壁模式反馈控制研究

IF 2.1 4区工程技术

Journal of Fusion Energy Pub Date : 2025-10-17 DOI: 10.1007/s10894-025-00519-x

Jing Wu, Shuo Wang, Xue-Feng Chen, Guang-Zhou Hao, Lei Xue, Yue-bin Liang, Peng Lu, Lie-Ming Yao

引用次数: 0

Tritium Management for LHD Deuterium Plasma Experiments LHD氘等离子体实验氚管理

IF 2.1 4区工程技术

Journal of Fusion Energy Pub Date : 2025-10-15 DOI: 10.1007/s10894-025-00518-y

Masahiro Tanaka, Hiromi Kato, Miki Nakada, Saori Kurita, Chie Iwata, Hiroki Chimura, Naoyuki Suzuki

{"title":"Tritium Management for LHD Deuterium Plasma Experiments","authors":"Masahiro Tanaka, Hiromi Kato, Miki Nakada, Saori Kurita, Chie Iwata, Hiroki Chimura, Naoyuki Suzuki","doi":"10.1007/s10894-025-00518-y","DOIUrl":"10.1007/s10894-025-00518-y","url":null,"abstract":"<div><p>In high-temperature plasma experiments using deuterium gas in large fusion devices, tritium is produced by the deuterium fusion reaction. Although the amount of tritium produced is not large, it is a radioactive material, so it is essential to develop safe handling systems and obtain public acceptance. Here, we summarize tritium safety management and tritium behavior in the facility, and have monitored the results of six years of deuterium plasma experiments in the Large Helical Device (LHD). More than 95% of the tritium exhausted from the LHD vacuum vessel was recovered by a tritium removal system. The extremely low concentrations of tritium discharged from the stack into the environment were monitored for each chemical form of tritiated water vapor, tritiated molecular hydrogen, and tritiated hydrocarbons, and were verified to be well below the levels specified by management at the National Institute for Fusion Science (NIFS). Although the tritium handled in the LHD deuterium experiment was a small amount, the operational experience and the instruments thereby developed would likely be useful for tritium safety management in future fusion reactors.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"44 2","pages":""},"PeriodicalIF":2.1,"publicationDate":"2025-10-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://link.springer.com/content/pdf/10.1007/s10894-025-00518-y.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145315852","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Recent Developments of the in-divertor Optical Box of the ITER Erosion Deposition Monitor ITER侵蚀沉积监测仪内导光箱的最新进展

IF 2.1 4区工程技术

Journal of Fusion Energy Pub Date : 2025-10-14 DOI: 10.1007/s10894-025-00517-z

Miklos Palankai, Teteny Baross, Jeno Kadi, Tamas Turcsik, Matyas Toth, Eszter Szucs, Gabor Szarvas, Fruzsina Daranyi, Attila Bohm, Gabor Veres, Mark Kempenaars, Govindarajan Jagannathan

{"title":"Recent Developments of the in-divertor Optical Box of the ITER Erosion Deposition Monitor","authors":"Miklos Palankai, Teteny Baross, Jeno Kadi, Tamas Turcsik, Matyas Toth, Eszter Szucs, Gabor Szarvas, Fruzsina Daranyi, Attila Bohm, Gabor Veres, Mark Kempenaars, Govindarajan Jagannathan","doi":"10.1007/s10894-025-00517-z","DOIUrl":"10.1007/s10894-025-00517-z","url":null,"abstract":"<div><p>The primary purpose of the ITER Erosion Deposition Monitor (EDM) is to track the erosion and deposition conditions of the Divertor Vertical Targets, as well as to monitor any changes in topology and surface damage resulting from plasma-wall interactions at these targets. The optical box of this diagnostic system, which is mounted on one of the Divertor Cassettes beneath the Divertor Dome, serves to provide a rigid support and protective shielding for the optical components housed inside. The EDM is an optical diagnostic system, whose accuracy depends on the integrity of the mechanical parts. The most critical in-vessel parts of this diagnostic system can be found on one of the Divertor Cassettes of the ITER machines, exposed to high heat and electromagnetic (EM) loads. The design challenges involve finding a solution for the proper heat transfer between the different components and besides, it is creating a solid design that can withstand the loads the optical box is exposed to. To achieve the required design level, the optical box and its components went through several design changes and simulation stages. This paper focuses on the challenges of the mechanical development of the parts that can be found on the Divertor Cassette. These components are the structural elements of the optical box and the support brackets of the mirrors inside the box. The paper presents the results of the analyses and the mechanical solutions of the structure of the box and the mirror supports which could withstand the high heat load and mechanical stresses coming from the EM loads.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"44 2","pages":""},"PeriodicalIF":2.1,"publicationDate":"2025-10-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://link.springer.com/content/pdf/10.1007/s10894-025-00517-z.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145315993","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Active Divertor Heat Flux Control using Impurity Powder Dropper 采用杂质粉末滴管的主动导流器热流控制

IF 2.1 4区工程技术

Journal of Fusion Energy Pub Date : 2025-10-06 DOI: 10.1007/s10894-025-00513-3

M. Ono, R. Raman, R. Maingi, S. Kaye, A. Sanchez-Villar

{"title":"Active Divertor Heat Flux Control using Impurity Powder Dropper","authors":"M. Ono, R. Raman, R. Maingi, S. Kaye, A. Sanchez-Villar","doi":"10.1007/s10894-025-00513-3","DOIUrl":"10.1007/s10894-025-00513-3","url":null,"abstract":"<div><p>Divertor plasma-facing components (PFCs) in a tokamak are typically designed to withstand average steady-state heat loads of about 5–10 MW/m², a limit that applies to both solid and liquid lithium (LL) PFCs. Exceeding these design values can result in surface damage to tungsten PFCs or excessive lithium (Li) evaporation in liquid lithium divertor (LLD) PFCs. Since exceeding the divertor heat load limits has serious consequences, it is therefore prudent to develop a tool to reduce the divertor heat load and bring the heat load to within the design limit without affecting the plasma performance. Active low Z impurity injection such as Li has been suggested as a potential solution to mitigate excess heat flux as suggested previously, given that non-coronal radiation can be quite large ~ 20–30 MJ per mole of injected Li. Li is considered desirable for reducing the edge neutral recycling helping to improve plasma energy confinement. In this paper, we model the Impurity Power Dropper (IPD) to investigate its potential of divertor heat flux control. The IPD is typically located at the top of the tokamak device and uses a vertical drift tube of a few meters. In the 2 m drift tube case, the IPD powder is accelerated to ~ 6 m/sec before reaching the plasma with the upper divertor configuration, matching the condition for the in-board side pellet injection case. By modeling the IPD geometry we determined the IPD powder deposition profile, and thus the non-coronal radiation and ionization profiles in time as well. From the enhanced radiation power loss, it is therefore possible to reduce the divertor heat load using the divertor simulation code. The IPD divertor heat flux control can be tested in the facilities with IPD including ST-40, DIII-D, EAST, WEST and NSTX-U.</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"44 2","pages":""},"PeriodicalIF":2.1,"publicationDate":"2025-10-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145256228","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"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

Spectroscopic Diagnostics of Non-Maxwellian Electron Velocity Distribution Function in the Large Helical Device 大型螺旋装置中非麦克斯韦电子速度分布函数的光谱诊断

IF 2.1 4区工程技术

Journal of Fusion Energy Pub Date : 2025-09-27 DOI: 10.1007/s10894-025-00516-0

Tomoko Kawate, Motoshi Goto

引用次数: 0

Developing Integrated Cost Models for Fusion Power Plants 发展核聚变电厂的综合成本模型

IF 2.1 4区工程技术

Journal of Fusion Energy Pub Date : 2025-09-24 DOI: 10.1007/s10894-025-00515-1

Rhian Chapman

{"title":"Developing Integrated Cost Models for Fusion Power Plants","authors":"Rhian Chapman","doi":"10.1007/s10894-025-00515-1","DOIUrl":"10.1007/s10894-025-00515-1","url":null,"abstract":"<div><p>Systems models and associated cost analyses are widely used within the fusion community to analyse tokamak designs, from prototype and demonstrator machines to potential commercial fusion power plants. To ensure the design programmes of fusion prototype/demonstrator power plants deliver a cost optimised design (within existing uncertainty limitations) the use of integrated cost modelling during the design process is essential. This integration produces holistic solutions in which engineering design choices and changes are directly represented in the cost results, allowing alternative solutions to be tested technologically and financially in the same analysis and cost estimates to be directly aligned with each specific design solution. Using examples from the STEP (Spherical Tokamak for Energy Production) programme this paper shows how implementing such an approach allows an interrogation of the design through the lens of cost-effectiveness, enabling a systematic exploration of potential trade-offs between performance and cost, highlighting cost drivers and interrogating the design aspects underpinning them, and facilitating holistic comparisons between design options. Including cost analysis into early design decisions through integrated cost modelling will drive a cost-optimised design; this is vital in proving that fusion power plants can be an economically viable energy source. One top-level example of this approach is understanding the critical size drivers and therefore cost drivers of the design, such as the inboard radial build for the STEP design. This understanding enables optimisation of this parameter within the relevant margins required to ensure performance (within design uncertainties).</p></div>","PeriodicalId":634,"journal":{"name":"Journal of Fusion Energy","volume":"44 2","pages":""},"PeriodicalIF":2.1,"publicationDate":"2025-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://link.springer.com/content/pdf/10.1007/s10894-025-00515-1.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145144599","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0