Fusion Engineering and Design最新文献

{"title":"Design progress of an articulated robotic arm for low-payload maintenance tasks in KSTAR","authors":"Dohee Lee ,&nbsp;Hong-Tack Kim ,&nbsp;Young Min Park ,&nbsp;Kwon Hee Hong ,&nbsp;Nam Il Her ,&nbsp;Jungsup Choi ,&nbsp;Jinhyun Kim ,&nbsp;Beom Seok Kim ,&nbsp;Jeong Whan Moon ,&nbsp;Sung Moo Ryew","doi":"10.1016/j.fusengdes.2025.115343","DOIUrl":"10.1016/j.fusengdes.2025.115343","url":null,"abstract":"<div><div>Maintenance of fusion experimental devices like KSTAR is challenging due to harsh conditions such as high vacuum and temperatures. Typically, long downtime is required to cool and reduce radiation levels before possible human access. To address this, we designed an articulated robotic arm to perform low-payload maintenance tasks inside the KSTAR device.</div><div>This paper presents the design progress of an articulated robotic arm, including system configuration, hardware design, and structural analysis, as part of a stage-wise approach to developing the arm. The arm is connected to the shuttle device and stored in a cask, totaling 11 m and 13 DoF. We conduct a finite element method (FEM) analysis to ensure design reliability and safety under target load conditions. We use the proposed robot system to perform essential maintenance operations for KSTAR, including visual inspection, debris removal, and other simple tasks. Our robot arm system has the potential to reduce maintenance preparation time, minimize radiation exposure to personnel, and contribute to improving the fusion experiment’s operational uptime and efficiency.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"220 ","pages":"Article 115343"},"PeriodicalIF":1.9,"publicationDate":"2025-07-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144654004","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":"Irradiation responses of an oxide-dispersion strengthened 15-15Ti austenitic stainless steel after He and D ion irradiations","authors":"Aoxiang Gong ,&nbsp;Chunfeng Shi ,&nbsp;Chaojun Luo ,&nbsp;Lu Sun ,&nbsp;Chi Xu ,&nbsp;Zhenfeng Tong","doi":"10.1016/j.fusengdes.2025.115336","DOIUrl":"10.1016/j.fusengdes.2025.115336","url":null,"abstract":"<div><div>Microstructure of 15-15Ti steel and oxide-dispersion strengthened (ODS) 15-15Ti steel samples after D/He ion irradiations was characterized by transmission electron microscopy (TEM). Both materials exhibit high densities of dislocation loops following He irradiation, with Frank dislocation loops (FDL) in edge-on geometry clearly resolved via rel-rod dark-field technique. Concomitantly, high densities of He bubbles were also observed in both samples. Radiation defects such as the \"black-dot\" defects were generated under D ion irradiation in both samples. The nanoindentation results indicate that irradiation hardening occurred in both samples, and it is shown that bubbles are weak strength obstacles, while dislocation loops are medium to strong strength obstacles. Radiation induced segregation (RIS) is observed at the grain boundaries of both samples after irradiation by the energy-dispersive spectrum analysis. Irradiation defects and irradiation effects observed for ODS samples are all lower than for 15-15Ti samples. The oxide particles stay stable in the irradiated ODS sample after different irradiation processes. The pinning of dislocation loops by oxide particles and the adsorption of He bubbles at the oxide particle/matrix interface can be observed, indicating the beneficial effects of the oxide particles in improving the irradiation resistance of the 15-15Ti steel.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"220 ","pages":"Article 115336"},"PeriodicalIF":1.9,"publicationDate":"2025-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144654001","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":"Experimental and numerical analysis of a D-shaped no-insulation high-temperature superconducting toroidal magnet for spherical Tokamaks","authors":"Zhengxin Yin ,&nbsp;Yi Tan ,&nbsp;Lang Qin ,&nbsp;Run Wu ,&nbsp;Yongqi Yang ,&nbsp;Lele Jiang ,&nbsp;Xiaofeng Wei ,&nbsp;Shouzhi Wang ,&nbsp;Binbin Wang ,&nbsp;Rui Chen","doi":"10.1016/j.fusengdes.2025.115338","DOIUrl":"10.1016/j.fusengdes.2025.115338","url":null,"abstract":"<div><div>The Spherical Tokamak (ST), with a low aspect ratio, improves plasma confinement and achieves a high <em>β</em> (the ratio of particle pressure to magnetic field pressure), thereby reducing the cost of the Toroidal Field (TF) magnet. However, the limited space in the center column of the ST necessitates a high engineering current density. Therefore, the no-insulation (NI) high-temperature superconducting (HTS) coil, known for its self-protection and high current-carrying capability, is inferred to be one of the competitive TF magnet technologies for future fusion power plants. Due to the ultra-compact geometry, the NI HTS TF coil used in Spherical Tokamaks differs from the constant tension <span>d</span>-shaped coil used in conventional Tokamaks, featuring a sharper <span>d</span>-shape and a smaller radius of curvature. Additionally, its current distribution characteristics, over-current performance, and multiple cooling cycle performance remain unclear. This study establishes an equivalent circuit model for the TF coil, demonstrating the current distribution inside the TF coil during charging, discharging, and over-current conditions. A laboratory-scale TF coil one meter in height, was manufactured to validate the model. Charging/discharging, over-current, and cooling cycle tests were conducted to evaluate the performance of the NI TF coil. Throughout the entire testing process, the voltage and magnetic field signals from both experiments and simulations were highly consistent. The model effectively calculates the toroidal and radial current distributions in the NI HTS TF coil, providing valuable insights into the design to prevent quenching. After two over-current tests and four cooling cycle tests, the coil's inductance and turn-to-turn resistance showed minimal change. This demonstrates the NI HTS coil's feasibility for application in next-generation experimental Spherical Tokamaks by Startorus Fusion.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"220 ","pages":"Article 115338"},"PeriodicalIF":1.9,"publicationDate":"2025-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144654002","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":"Reducing laser plasma instability in inertial confinement fusion by polarization smoothing based on stress birefringence","authors":"Bowu Liu ,&nbsp;Yu Xie ,&nbsp;Hui Wang ,&nbsp;Qiang Yuan ,&nbsp;Xibo Sun ,&nbsp;Honghong Wu ,&nbsp;Hui Zhang","doi":"10.1016/j.fusengdes.2025.115337","DOIUrl":"10.1016/j.fusengdes.2025.115337","url":null,"abstract":"<div><div>Polarization smoothing is one of the effective means to reduce laser plasma instability (LPI) in inertial confinement fusion (ICF). However, with the increase in laser power, there is currently a lack of stable single-beam polarization smoothing methods suitable for engineering applications. In our previous work, we proposed a polarization smoothing idea based on stress birefringence of edge loading fused silica window, but two issues were left unresolved: determination of the optimal edge loads distribution and design of the loading structure within limited available space. In this article, a method for determining the optimal loads combining continuous load discretization processing and genetic algorithm is proposed. And guided by the optimized results, we propose a pre-tightening loading method based on the reaction force of deflection deformation of special-shaped beams. Enabling the loading structure to provide the optimal edge loads for the stressed optical window while being thin enough to allow the element to be installed into ICF final optics assembly. We conducted tests on the element in the ultraviolet (351 nm wavelength) section of a high-power laser system, with a laser energy of 2kJ/3 ns. The results showed that compared to the situation without polarization smoothing, the backscattered energy decreased by 50∼60%, proving the effectiveness of this polarization smoothing scheme in reducing LPI.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"220 ","pages":"Article 115337"},"PeriodicalIF":1.9,"publicationDate":"2025-07-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144633003","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":"Modelling and design of magnetic coil power supply for horizontal displacement feedback field of TT-1 tokamak","authors":"Yu Zhou ,&nbsp;Zhenyu Li ,&nbsp;Yiyun Huang","doi":"10.1016/j.fusengdes.2025.115335","DOIUrl":"10.1016/j.fusengdes.2025.115335","url":null,"abstract":"<div><div>The horizontal displacement feedback field(HDFF) magnet power supply, is a significant subsystem of magnet power supply system for Thailand Tokamak Device 1(TT-1). To achieve feedback adjustment of plasma horizontal position, the coil current needs to meet a response time of 1 ms and a steady-state performance higher than 1 %. Therefore, interleaved parallel buck converters are adopted as the power supply current regulation method. To design a suitable control system to achieve the output performance of the power supply, the state-space smoothing method was used to establish the transfer function of the system. And verify its accuracy through circuit simulation. With the help of the current duty cycle transfer function and a suitable PI controller, the output current reaches 230A within 1mS, and the steady-state error coefficient and ripple coefficient are also below 1 %. The test results demonstrate that the HDFF power supply can meet the system requirements, and it has strong reference significance, which can provide reference for the design of other fast and high-precision magnetic field power supply in magnet systems.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"220 ","pages":"Article 115335"},"PeriodicalIF":1.9,"publicationDate":"2025-07-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144654003","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":"Investigation of laser powder bed fusion for the application of additive manufacturing in the First Wall of the DTT facility","authors":"G. De Sano ,&nbsp;M. Furno Palumbo ,&nbsp;R. De Luca ,&nbsp;G. Dose ,&nbsp;M. Gelfusa ,&nbsp;A. Pizzuto ,&nbsp;G.M. Polli ,&nbsp;B. Riccardi ,&nbsp;S. Roccella","doi":"10.1016/j.fusengdes.2025.115334","DOIUrl":"10.1016/j.fusengdes.2025.115334","url":null,"abstract":"<div><div>The design of plasma-facing components for fusion devices is a challenging task, due to the requirements of high geometrical precision of complex parts, as well as the need of optimizing the thermo-structural performance of the adopted materials, which are required to sustain significant heat and electro-magnetic loads. This issue has been tackled in the Divertor Tokamak Test (DTT) facility within the manufacturing activities of the First Wall (FW), where the potential application of advanced technologies to fabricate structural parts made of 316L stainless steel is being investigated. In this framework, laser powder bed fusion of metals (PBF-LB/M) has been identified as a promising additive manufacturing solution, offering unique advantages in terms of design flexibility and performance of the produced material.</div><div>This study presents an examination of the preliminary results from the fabrication and test of small samples and mock-ups, with the goal of performing a first qualification of the technology for our application. The preparatory samples allowed us to evaluate the capability of PBF-LB/M to produce low-porosity parts with enhanced mechanical properties, also including a thermal treatment for stress relief. On the other hand, compatibility with the design requirements has been assessed directly on the mock-ups, which are representative of the FW layout. The results obtained confirmed the advantages of PBF-LB/M for its employment in the manufacturing route of the DTT FW, while the encountered limitations were not identified as a showstopper. Therefore, the findings of this research contribute both to the preparation activities of the DTT FW and to the advancement of additive manufacturing technologies for nuclear fusion applications.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"220 ","pages":"Article 115334"},"PeriodicalIF":1.9,"publicationDate":"2025-07-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144631762","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":"Nanomechanical property and microstructural changes of tungsten Langmuir probes retrieved from JET with metal walls","authors":"E. Fortuna-Zalesna ,&nbsp;M. Spychalski ,&nbsp;P. Petersson ,&nbsp;M. Rubel ,&nbsp;A. Widdowson ,&nbsp;B. Thomas ,&nbsp;D. Moszczyńska ,&nbsp;P. Wiecinski ,&nbsp;Ł. Ciupiński ,&nbsp;J. Likonen ,&nbsp;R. Kerr ,&nbsp;JET Contributors","doi":"10.1016/j.fusengdes.2025.115319","DOIUrl":"10.1016/j.fusengdes.2025.115319","url":null,"abstract":"<div><div>Three tungsten Langmuir probes retrieved from the JET tokamak with the ITER-Like Wall (JET-ILW) from the bulk tungsten Tile 5 have been studied. Nano-indentation, microscopy, ion beam analysis (IBA), and X-ray diffraction were used to assess changes in their mechanical properties, microstructure, and phase composition. Four regions of the probes were studied - the tip and the base, at two sides: front and back. The hardness value of one of the probes (no. 5, Stack B) in the tip area was reduced when compared to the value measured on the base section: 5.4 GPa versus 8.8 GPa, respectively. On the two other probes, the hardness was similar to that of the reference material. At the protrusion of probe 5, the recrystallized zone was observed. The IBA analysis revealed that the probes’ surfaces below the tips were covered by a thin layer of deposit composed primarily of beryllium, oxygen, carbon, and hydrogen isotopes, along with smaller amounts of nickel, nitrogen, and helium at some locations. The presence of tungsten carbide W₂C was revealed on the tip of probe 5, in the area where IBA measurements indicated elevated carbon content in the material, demonstrated by analysis of the XRD records.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"220 ","pages":"Article 115319"},"PeriodicalIF":1.9,"publicationDate":"2025-07-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144614874","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":"On the minimum size of tokamaks aimed at fusion energy gain with optimum plasma temperature","authors":"J.R. Wu ,&nbsp;G.S. Xu ,&nbsp;D.H. Chen ,&nbsp;F.J. Wang ,&nbsp;B.N. Wan","doi":"10.1016/j.fusengdes.2025.115308","DOIUrl":"10.1016/j.fusengdes.2025.115308","url":null,"abstract":"<div><div>This work explores factors affecting the size reduction of tokamaks aiming at a specific thermal deuterium-tritium fusion energy gain <span><math><msub><mi>Q</mi><mrow><mi>f</mi><mi>u</mi><mi>s</mi></mrow></msub></math></span>, using 1/2D physics models with IPB98(y,2) scaling law. A simplified <span><math><msub><mi>Q</mi><mrow><mi>f</mi><mi>u</mi><mi>s</mi></mrow></msub></math></span> expression reveals an optimum plasma temperature that minimizes device size for a given <span><math><msub><mi>Q</mi><mrow><mi>f</mi><mi>u</mi><mi>s</mi></mrow></msub></math></span>. Conversely, this temperature maximizes <span><math><msub><mi>Q</mi><mrow><mi>f</mi><mi>u</mi><mi>s</mi></mrow></msub></math></span> for fixed device dimensions. Parametric explorations show that optimum temperatures depend primarily on <span><math><msub><mi>Q</mi><mrow><mi>f</mi><mi>u</mi><mi>s</mi></mrow></msub></math></span>, plasma profiles and impurity fraction, with negligible sensitivity to parameters like edge safety factor <span><math><msub><mi>q</mi><mn>95</mn></msub></math></span>, energy confinement factor <span><math><msub><mi>H</mi><mrow><mn>98</mn><mi>y</mi><mn>2</mn></mrow></msub></math></span>, radial build dimensions and so on. Based on the results of <span><math><msub><mi>Q</mi><mrow><mi>f</mi><mi>u</mi><mi>s</mi></mrow></msub></math></span> expression and parametric explorations, smaller device sizes require assumptions of optimum temperature, higher Greenwald fraction <span><math><msub><mi>f</mi><mrow><mi>G</mi><mi>W</mi></mrow></msub></math></span>, <span><math><msub><mi>H</mi><mrow><mn>98</mn><mi>y</mi><mn>2</mn></mrow></msub></math></span>, peaking factors of plasma profiles, lower <span><math><msub><mi>q</mi><mn>95</mn></msub></math></span> and impurity fraction. Further, the design points at optimum temperatures are evaluated against the constraints including neutron wall loading, divertor heat load and pulse length. In addition to plasma temperature, the device size dependence on maximum toroidal field <span><math><msub><mi>B</mi><mrow><mi>m</mi><mi>a</mi><mi>x</mi></mrow></msub></math></span> is also found to be non-monotonic. On one hand, higher magnetic field reduces the required plasma minor radius to achieve a certain <span><math><msub><mi>Q</mi><mrow><mi>f</mi><mi>u</mi><mi>s</mi></mrow></msub></math></span>. On the other hand, higher magnetic field increases structural demands on toroidal field (TF) coils to alleviate the magnetic stress and sustain the high field. As a result, the competition between both effects leads to a decrease and then an increase of the whole device size as magnetic field increases.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"220 ","pages":"Article 115308"},"PeriodicalIF":1.9,"publicationDate":"2025-07-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144605503","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":"Tests of power supply and fast discharge unit for the Frascati Coil Cold Test Facility","authors":"A. Lampasi ,&nbsp;M. Manganelli ,&nbsp;S. Pipolo ,&nbsp;R. Romano ,&nbsp;B. Taheri ,&nbsp;P. Zito","doi":"10.1016/j.fusengdes.2025.115310","DOIUrl":"10.1016/j.fusengdes.2025.115310","url":null,"abstract":"<div><div>Cold test facilities are essential for the advancement of fusion energy research. The Frascati Coil Cold Test Facility (FCCTF) was envisioned to test DTT superconducting coils at cryogenic temperature and full current. The FCCTF is also useful for early optimization and validation of other key technologies such as power supply (PS), quench detection system and protection fast discharge unit (FDU).</div><div>The PS is rated 44 kA with 24-pulse topology to meet the requirement for low current ripple even for a single load coil. The FDU design is based on a fully static breaking mechanism and on nonlinear varistors for energy dissipation. The FDU can be configured to comply with all possible requirements of central solenoid, toroidal field and poloidal field coils in terms of current, peak voltage, discharge time and energy dissipation.</div><div>This paper presents the FCCTF design and the most relevant acceptance tests performed on its PS and FDU, showing that they are suitable for their purpose.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"220 ","pages":"Article 115310"},"PeriodicalIF":1.9,"publicationDate":"2025-07-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144611605","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":"Lingshu: A modular real-time control system framework for experimental tokamaks","authors":"Z.M. Huang ,&nbsp;Q.P. Yuan ,&nbsp;J.J. Huang ,&nbsp;J.Q. Zhu ,&nbsp;G. Xu ,&nbsp;R.R. Zhang ,&nbsp;B.J. Xiao","doi":"10.1016/j.fusengdes.2025.115309","DOIUrl":"10.1016/j.fusengdes.2025.115309","url":null,"abstract":"<div><div>The next generation of Tokamak devices in China necessitates enhanced real-time and scalable control systems. A new control system framework, Lingshu, has been designed and implemented to construct a comprehensive Plasma Control System (PCS). Lingshu employs a modular architecture based on a lightweight core infrastructure to achieve flexibility while maintaining real-time performance. The system is structured into three layers according to the tendency to be varied. The core layer constructs the framework fundamentals with built-in functionalities, including execution sequencing, data management, and communication mechanisms. The component suite layer determines the control system form by a set of collaborating components. The control algorithm layer provides the system with control capabilities. This three-tier architecture creates clear boundaries within the system. It avoids losing flexibility due to high integration or complexity booming due to excessive modularity. Lingshu employs an approach to modeling data elements to standardize the integration of algorithms. This approach allows automated tools to extract information from the model and utilizing, thereby streamlining the development and integration of new algorithms. This work explores innovative practices in designing and engineering the plasma control system framework and has achieved positive application results. The first release of Lingshu has been adapted for the Experimental Advanced Superconducting Tokamak (EAST). It incorporates all requisite control functions for routine operations on EAST and has been validated in over 1,000 discharge shots conducted during 2023 and 2024. Experimental validation on EAST demonstrates its reliability, scalability, and ability to maintain stable plasma control across multiple discharge scenarios.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"220 ","pages":"Article 115309"},"PeriodicalIF":1.9,"publicationDate":"2025-07-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144604986","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}