Fusion Engineering and Design最新文献

{"title":"The control and data acquisition system of the DTT experiment","authors":"G. Manduchi ,&nbsp;F. Zanon ,&nbsp;L. Boncagni ,&nbsp;P. Mosetti ,&nbsp;G. Martini ,&nbsp;G. Paccagnella ,&nbsp;C. Centioli ,&nbsp;R. Ambrosino ,&nbsp;F. Sartori","doi":"10.1016/j.fusengdes.2024.114777","DOIUrl":"10.1016/j.fusengdes.2024.114777","url":null,"abstract":"<div><div>DTT, Divertor Tokamak Test facility, is currently under construction at the Frascati ENEA Research Center. Its main aim is to explore alternative solutions for the extraction of the heat generated by the fusion process. Its Control and Data Acquisition System (CODAS) will (1) orchestrate and synchronize all the DTT systems during Plasma operation and maintenance; (2) acquire data from the experiment diagnostics and plant systems and store it in an experimental database to be used for on-line and off-line analysis; (3) provide real-time Plasma control. The expected duration of the plasma discharge in DTT is in the order of some tens of seconds and therefore DTT can be considered a long-lasting experiment, involving data streaming technologies for data communication and storage. The main DTT CODAS design is based on three principles: (1) Taking inspiration from other similar experiments currently under development, namely ITER CODAC, (2) relying on proven solutions already adopted in running experiments with similar constraints and (3) taking advantage from practices widely adopted in fusion and, more in general, in industry. Despite using components already adopted in other fusion experiments, DTT CODAS is the first system that seamlessly integrates all of them.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"211 ","pages":"Article 114777"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143137270","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":"Study on the changes in the reverse recovery characteristics of high-power thyristor under 14.1 MeV fusion neutron irradiation","authors":"Wei Tong ,&nbsp;Hua Li ,&nbsp;Dongmei Liu ,&nbsp;Yanan Wu ,&nbsp;Meng Xu ,&nbsp;Kun Wang","doi":"10.1016/j.fusengdes.2024.114744","DOIUrl":"10.1016/j.fusengdes.2024.114744","url":null,"abstract":"<div><div>When the <span>d</span>-T fusion reaction takes place in a Tokamak, the high-energy neutrons emitted, with an energy of 14.1 MeV, can alter the electrical properties of high-power thyristors in Quench Protection System (QPS). The Reverse Recovery Characteristics (RRC) of High-power Thyristor (HP-SCR) is one of the crucial issues affecting the reliability of QPS. In this paper, the change in the RRC of HP-SCR under 14.1 MeV neutron irradiation is deeply studied. Firstly, the microscopic material damage mechanism of HP-SCR induced by neutron irradiation and its relationship with RRC are deeply analyzed. Secondly, a highly efficient neutron irradiation experiment is designed and conducted to effectively validate the correctness of the theoretical analysis regarding changes in RRC. Finally, the effects of neutron irradiation on the QPS, consisting of multiple thyristors arranged in series, are analyzed and discussed through system-level simulations. The study offers important recommendations for the maintenance and upgrade of QPS, which will greatly enhance the safety of Tokamak devices.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"211 ","pages":"Article 114744"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143136889","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":"High heat flux tests of tungsten – RAFM steel mock-up brazed by Ti-Zr-4Be alloy","authors":"D.M. Bachurina ,&nbsp;A.N. Suchkov ,&nbsp;O.N. Sevryukov ,&nbsp;V.O. Kirillova ,&nbsp;M.V. Leontyeva-Smirnova ,&nbsp;T.N. Vershinina ,&nbsp;P.Yu. Piskarev ,&nbsp;V.V. Ruzanov ,&nbsp;V.E. Kuznetcov ,&nbsp;M.S. Kolesnik ,&nbsp;I.V. Mazul","doi":"10.1016/j.fusengdes.2024.114790","DOIUrl":"10.1016/j.fusengdes.2024.114790","url":null,"abstract":"<div><div>This work focuses on the development of Plasma Facing Components (PFC) for future fusion reactors. One of the main challenges is to provide a firm joint between Plasma Facing Materials (PFM), such as tungsten (W), and structural material, such as Reduced-Activation Ferritic-Martensitic (RAFM) steel. This work advances the development of the W/48Ti-48Zr-4Be/Ta/48Ti-48Zr-4Be/Rusfer joint, where 48Ti-48Zr-4Be wt.% is a brazing alloy and Ta is an interlayer that mitigates the mismatch in the CTE between W and the Rusfer steel. We investigated the high heat flux performance (HHFT) of this joint, which included the design of a water-cooled PFC, its fabrication and HHFT at the IDTF stand. A thermal cycling test with a stepwise increase in heat flux from 2.5 MW/m² to 4.5 MW/m² was conducted. The breakdown of the mock-up took place after 1055 cycles at the Ta/RAFM surface. The fracture surface was investigated using SEM and XRD, which revealed fatigue failure along the (Zr,Ti)₂(Fe,Cr,Ta) and Ta₂Be phases.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"211 ","pages":"Article 114790"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143136993","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":"A structural optimized method and verification of fusion converter for high current sharing performance","authors":"Zhongma Wang ,&nbsp;Xingxing Shao ,&nbsp;Jiameng Wu ,&nbsp;Tao Xu ,&nbsp;Bowen Chu ,&nbsp;Shenao Zhang ,&nbsp;Yujie Lu ,&nbsp;Chuangye Yin ,&nbsp;Wenwu Lu ,&nbsp;Sheng Zhang ,&nbsp;Xiuqing Zhang ,&nbsp;Yanchun Wang","doi":"10.1016/j.fusengdes.2024.114793","DOIUrl":"10.1016/j.fusengdes.2024.114793","url":null,"abstract":"<div><div>Current unbalance caused by impedance disparities among multiple parallel branches can lead to accelerated aging of some power devices and localized overheating, posing a threat to the normal operation of fusion power supplies. A method for improving the current sharing performance of fusion converters with paralleled branches based on structural parameter optimization is proposed in this paper. Firstly, a high-precision converter bridge arm structure is constructed using ANSYS Q3D, and sub-circuits of segmented modules are extracted; subsequently, a circuit model is built in Simplorer, and the current of each branch is obtained through joint simulation with Simplorer; then, with branch current as the matching parameter, an adaptive target optimization algorithm is utilized to iterate and optimize structural parameters, thereby acquiring a set of bridge arm structures with optimal current sharing performance; finally, fine-tuning is conducted considering the actual spatial limitation. To validate the superiority of the proposed method, the current sharing effects of the structures before and after optimization under both steady-state and short-circuit conditions are analyzed and compared. The results indicate that the optimized structure has shown significant improvements in terms of current sharing coefficient, over-current ratio, and maximum turn-off error time. This method could be used for the current sharing design of fusion power supplies and related multiple parallel devices.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"211 ","pages":"Article 114793"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143136996","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":"MDSplus version 8.0 – The path to long node names","authors":"Mark Winkel,&nbsp;Joshua Stillerman,&nbsp;Stephen Lane-Walsh,&nbsp;Fernando Santoro","doi":"10.1016/j.fusengdes.2024.114742","DOIUrl":"10.1016/j.fusengdes.2024.114742","url":null,"abstract":"<div><div>The MDSplus open-source software used to record, visualize and archive fusion data, has always been limited to 12-character node names. That is insufficient to support descriptive names, such as those used in ITER's IMAS data schema. A prototype of MDSplus v8.0 is described that expands node names to 63 characters. Discussion covers additional proposed features, internal implementation details and tips on migrating to v8.0 when it is eventually released. MDSplus v8.0 uses a different file format to accommodate the long node names, thus it is a “breaking” change. Features to facilitate machine learning applications are also being considered. MDSplus v8.0 and v7.x are essentially different products; customers must choose which product they wish to use. The feature set of MDSplus v8.0 is still being defined thus MDSplus users are encouraged to submit their suggestions.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"211 ","pages":"Article 114742"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143137106","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":"Simulation analysis of the motion accuracy of the ECRH launcher steering mechanism based on ADAMS","authors":"Xiaoguang Zhang ,&nbsp;Nengtao Zhou ,&nbsp;Haibiao Zhang ,&nbsp;Dingzhen Li ,&nbsp;Yu Hua ,&nbsp;Chao Xu ,&nbsp;Chao Zhang ,&nbsp;Ming Gao ,&nbsp;Xiaojie Wang ,&nbsp;Yunying Tang","doi":"10.1016/j.fusengdes.2024.114774","DOIUrl":"10.1016/j.fusengdes.2024.114774","url":null,"abstract":"<div><div>The Electron Cyclotron Resonance Heating (ECRH) system is designed to inject radiofrequency power into the plasma for heating and current drive applications. The launcher steering mechanism is one of the critical components of the ECRH system. Therefore, this paper conducts an in-depth study of the launcher steering mechanism. The paper initially presents the structural forms of the ECRH launcher steering mechanism. Subsequently, it deduces the relational expression between the displacement of the drive rod and the rotation angle of the steerable mirror. Finally, utilizing the simulation analysis functionality of the virtual prototype software ADAMS, it examines the impact of rod length manufacturing errors and clearances in the rotary joint on the motion accuracy of the launcher steering mechanism, providing a basis for the manufacturing and assembly of the mechanism.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"211 ","pages":"Article 114774"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143136862","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":"Achieving effective bonding between W-75Cu composite and CuCrZr alloy via spark plasma sintering","authors":"Changcheng Sang ,&nbsp;Dang Xu ,&nbsp;Kaichao Fu ,&nbsp;Ruizhi Chen ,&nbsp;Pengqi Chen ,&nbsp;Yingwei Lu ,&nbsp;Yonghong Xia ,&nbsp;Qiu Xu ,&nbsp;Jigui Cheng","doi":"10.1016/j.fusengdes.2024.114791","DOIUrl":"10.1016/j.fusengdes.2024.114791","url":null,"abstract":"<div><div>In this study, spark plasma sintering (SPS) was employed to achieve the simultaneous sintering and bonding of W-75Cu composite with CuCrZr alloy. The effects of the sintering temperature on the microstructure evolution and properties of the W-75Cu/CuCrZr joints were systematically investigated and their thermal shock resistance was evaluated. The results indicated that a dense and defect-free joint was obtained at 950 °C, demonstrating the maximum shear strength (216.5 MPa) and thermal conductivity (237.9 W/(m·K)). Fracture analysis revealed that failure predominantly occurred within the W-75Cu matrix, confirming robust interfacial bonding. Additionally, after 200 thermal shock cycles at 450 °C-RT, the W-75Cu/CuCrZr joint maintained a high joint strength (172.7 MPa) without visible cracks on the interface, thereby demonstrating excellent joint reliability and thermal shock resistance. This study highlights the advantages of SPS technology in promoting the densification of the matrices and achieving high-performance joints, providing valuable technical insights for achieving a reliable bonding between the W-Cu FGM (high Cu content) and the CuCrZr heat sink material.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"211 ","pages":"Article 114791"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143136947","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":"Microstructural and mechanical analyses of YBCO coated conductor tapes in high-irradiation environments","authors":"Yuanzhou Pan ,&nbsp;Canjie Xin ,&nbsp;Wei Wu ,&nbsp;Mingzhi Guan ,&nbsp;Xingzhe Wang","doi":"10.1016/j.fusengdes.2025.114802","DOIUrl":"10.1016/j.fusengdes.2025.114802","url":null,"abstract":"<div><div>Upgrading the superconducting magnets in the High Energy Fragment Separator (HFRS) at the High Intensity Heavy Ion Accelerator Facility (HIAF) is challenging due to the complex irradiation field. YBCO coated conductor (CC) tapes in magnets are designed to resist high radiation heat loads and endure significant irradiation doses. However, prolonged operation will also result in degradation of the tapes. To assess the reliability of the superconducting magnets, it's crucial to study the microstructures and macro-behaviors of YBCO conductors subjected to higher irradiation doses. In this work, irradiation experiments were conducted using 160 MeV ⁴⁰Ar ions on YBCO conductors with/without Cu stabilizer, ranging from 4.8 × 10⁷ ions/cm² to 4.8 × 10¹² ions/cm². Micro-defects morphology analysis employed scanning electron microscopy (SEM), energy-dispersive x-ray spectroscopy (EDS) and transmission electron microscopy (TEM). Macro-measurements of unirradiated and irradiated YBCO conductors were conducted to investigate the irradiation dose dependence of superconducting properties and mechanical behaviors. Post high-dose irradiation, TEM and EDS analyses revealed YBCO CC tapes without Cu stabilizer exhibited more obvious delamination behaviors at the edge of YBCO layer due to the irradiation defects. Macro-behavior measurements indicated nonlinear dependencies of mechanical properties, critiical current (<em>I_c</em>) and critical tenperature (<em>T</em>_<em>c</em>) on irradiation fluences. Furthermore, for the purpose of comparison, tapes with Cu stabilizer demonstrated superior radiation-resistant properties than those without Cu stabilizer, and the mechanical parameters and superconducting behaviors of YBCO CC tapes without Cu stabilizer significantly degraded at high irradiation dose, thus, the irradiation damages in YBCO layer identified as the primary cause of <em>I_c</em> and <em>T</em>_<em>c</em> degradation under high irradiation fluence.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"211 ","pages":"Article 114802"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143136952","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":"Corrigendum to “Thermal stability of differently rolled pure tungsten plates in the temperature range from 1125°C to 1250°C” [Fusion Engineering and Design 192 (2023) 113581:1-6]","authors":"Thomas Larsen ,&nbsp;Kevin Chmelar ,&nbsp;Benjamin Lindegren Larsen ,&nbsp;Patrick Nagy ,&nbsp;Kang Wang ,&nbsp;Wolfgang Pantleon","doi":"10.1016/j.fusengdes.2025.114818","DOIUrl":"10.1016/j.fusengdes.2025.114818","url":null,"abstract":"","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"211 ","pages":"Article 114818"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143137522","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":"Physics and technology maturity level required for the K-DEMO design points","authors":"J.S. Kang ,&nbsp;G. Jo ,&nbsp;J-M. Kwon ,&nbsp;B.G. Hong","doi":"10.1016/j.fusengdes.2024.114741","DOIUrl":"10.1016/j.fusengdes.2024.114741","url":null,"abstract":"<div><div>The conceptual design study for the Korean fusion demonstration reactor (K-DEMO) has been conducted, placing an emphasis on the imperative to reduce the reactor's dimension as a means to cost minimization. The design space of K-DEMO was delineated based on maturity level of physics and technology. Advanced technology features such as a use of a tungsten carbide (WC) shield, a tritium breeding blanket concept of He cooled lithium lead (HCLL), the maximum allowable magnetic field at the TF coil, <em>B_max</em> = 16 T with a Nb₃Sn superconducting material, etc. were adopted. With specified design criteria, including a net electric power ≥ 300 MW, a fusion gain, <em>Q</em> &gt; 20.0, a neutron wall loading &lt; 2.0 MW/m², an indicator of divertor power handling capability (ratio of power to divertor to major radius), <em>P_div</em>/<em>R₀</em> &lt; 25 MW/m, and the capability for steady-state operation, a design space of the K-DEMO was established based on the energy confinement scaling law of IPB98[y,2] under physics level of Greenwald density fraction, <em>n_e</em>/<em>n_G</em> &lt; 1.2, normalized plasma beta (ratio of plasma pressure to magnetic pressure normalized by plasma current divided by the product of minor radius and toroidal magnetic field), <em>β_N</em> &lt; 3.0, confinement enhancement factor, <em>H</em> &lt; 1.3, and a direct cost ≤ 7.5 B$. After an exploration of system parameters, prospective design points for K-DEMO were identified, characterized by a major radius, <em>R₀</em> ∼ 6.8 m, an aspect ratio, <em>A</em> = 3.1, a toroidal magnetic field at plasma center, <em>B_T</em> ≥ 6.5 T, and a fusion power, <em>P_fusion</em> ∼ 1,500 MW. When <em>β</em>-independent energy confinement scaling law was applied, the design points were accessible with smaller <em>n_e</em>/<em>n_G, β_N, H, P_fusion</em>, and larger <em>B_T</em>. From a sensitivity analysis of the minimum major radius to the input parameters, strong sensitivities to Greenwald density fraction, <em>n_e/n_G</em>, normalized plasma beta, <em>β_N</em>, confinement enhancement factor, <em>H</em>, edge safety factor, <em>q_edge</em>, and elongation, <em>κ</em> were found. Additionally, the operational envelope in physics and technology parameters was established with system parameters associated with the design points.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"211 ","pages":"Article 114741"},"PeriodicalIF":1.9,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143136905","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}