Journal of Nuclear Materials最新文献

{"title":"Post-ballooning and burst steam oxidation of accident tolerant zirconium alloy cladding with cracked chromium coating","authors":"Hyunwoo Yook,&nbsp;Sunghoon Joung,&nbsp;Youho Lee","doi":"10.1016/j.jnucmat.2025.156095","DOIUrl":"10.1016/j.jnucmat.2025.156095","url":null,"abstract":"<div><div>This study presents a mechanistic model for coating cracking and subsequent oxidation of burst Cr-coated Zircaloy cladding under Design Basis Accident (DBA) conditions. A series of sequential simulated Loss Of Coolant Accident (LOCA) experiments were conducted using the <em>i</em>-LOCA facility and the high-temperature oxidation facility at Seoul National University. 1.5 m long Cr-coated claddings(15μm) with inserted ZrO₂ pellets were internally pressurized and inductively heated in an inert environment using the <em>i</em>-LOCA facility to induce ballooning and burst under various internal pressures and two pellet configurations—cylindrical and single powder pellet—designed to simulate unfragmented (&lt;55 GWd/MTU) and fully pulverized (∼94 GWd/MTU) fuel conditions, respectively. The resulting post-burst cladding geometries with various burst hoop strains were analyzed via 3D scanning. The burst region (±1.5 inches from the burst center) of the post-burst specimens was subsequently subjected to two-sided oxidation using the high-temperature oxidation facility.</div><div>Mechanistic models for coating cracking and Equivalent Cladding Reacted (ECR) calculation were developed based on the conventional definition of ECR and a phenomenological understanding derived from microstructural characterization of post-burst specimens, thereby providing a general framework applicable irrespective of the specific coating. Under DBA conditions, the additional oxidation attributable to coating cracking was quantified to be no more than 25 % relative to inner-sided oxidation. Accident coping time analyses indicated that, even when accounting for burst-induced deformation and coating cracking, the application of Cr coating maintains compliance with existing regulatory safety margins.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156095"},"PeriodicalIF":3.2,"publicationDate":"2025-08-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144842105","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Crystal plasticity model for UO2: Introduction of the dislocation climbing and coupling with the thermally activated gliding","authors":"S. ElBez ,&nbsp;M. Garajeu ,&nbsp;B. Michel","doi":"10.1016/j.jnucmat.2025.156083","DOIUrl":"10.1016/j.jnucmat.2025.156083","url":null,"abstract":"<div><div>Modelling of viscoplastic behaviour of <span><math><mrow><mi>U</mi><msub><mi>O</mi><mn>2</mn></msub></mrow></math></span> nuclear fuel at high temperature is of major interest to analyze the risk of cladding failure in power transient condition where pellet to cladding mechanical interaction occurs. In this study, we investigate the impact of dislocation climbing on the viscoplastic behaviour of <span><math><mrow><mi>U</mi><msub><mi>O</mi><mn>2</mn></msub></mrow></math></span> single crystal at high temperature where climbing mechanism is activated by vacancy diffusion. The modelling framework is that of an existing crystal plasticity law devoted to thermally activated gliding for lower temperatures. The constitutive equation of the crystal plasticity model are modified in order to add dislocation interaction hardening and stored dislocation recovery. The latter is assessed with dynamic recovery induced by cross slip and static recovery induced by dislocation climbing. A phenomenological model is proposed to compute the dislocation climbing velocity and the associated dislocation density recovery rate. The dislocation gliding velocity law is also improved with a physically based formulation introducing the activation energy for a double kink mechanism. The complete model, for low and high temperatures, is built with a harmonic coupling function enabling a more physically based formulation of the constitutive equations devoted to each mechanism. An implicit numerical implementation of the resulting model is proposed with a finite strain formulation in the framework of the MFront open source tool. A first validation of the complete model has been done with experiment/simulation comparisons for single crystal creep compression tests. In this first step, the validation is limited to orientations where only the modes 1 and 2 (soft and hard slip modes) are suspected of having a major contribution to the viscoplastic behaviour. Thanks to the contribution of dislocation gliding, interaction and climbing we can explain the evolution of the compression flow stress as a function of the temperature from 800 °C to 1600 °C in good agreement with experimental results. The physically based formulation gives also a justification of the temperature dependency of the apparent thermal activation energy and of the strain rate sensitivity.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156083"},"PeriodicalIF":3.2,"publicationDate":"2025-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144809876","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Mobility assessment of the BCC and carbide phases in the C-Nb, C-U and Nb-U systems","authors":"Nicholas Ury,&nbsp;Kyoung Kweon,&nbsp;Jibril Shittu,&nbsp;Aurélien P. Perron,&nbsp;Bradley C. Childs,&nbsp;Emily E. Moore","doi":"10.1016/j.jnucmat.2025.156084","DOIUrl":"10.1016/j.jnucmat.2025.156084","url":null,"abstract":"<div><div>Uranium carbides with refractory metal additions are considered for Gen IV nuclear reactors and nuclear thermal propulsion as fuels for their high-temperature and corrosion resistant properties. Understanding kinetic effects that dictate microstructural evolution during fabrication and operating conditions is essential to advance technological development of these fuels. This work presents the development of an atomic mobility database for C-Nb-U systems based off available experimental data supported with ab-initio methods. The mobility assessments and uncertainty quantification (using Markov chain Monte Carlo) were conducted in the Kawin software. Carbon diffusion is considered dominant, as metal diffusion is much slower, with niobium diffusion being even slower and rate limiting than uranium metal. We provide a comprehensive and self-consistent thermo-kinetic database that is validated by diffusion couple simulations through Kawin. This enables prediction of microstructural and phase evolution critical for the development and lifetime assessment of next generation nuclear fuels.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156084"},"PeriodicalIF":3.2,"publicationDate":"2025-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144810212","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"In-situ irradiation of uranium carbide","authors":"Rashed Almasri ,&nbsp;Adrian R. Wagner ,&nbsp;Laura Hawkins ,&nbsp;Wei-Ying Chen ,&nbsp;Jennifer K. Watkins ,&nbsp;Jian Gan ,&nbsp;Lingfeng He","doi":"10.1016/j.jnucmat.2025.156082","DOIUrl":"10.1016/j.jnucmat.2025.156082","url":null,"abstract":"<div><div>Uranium carbide (UC) is a leading candidate fuel for Generation IV reactors due to its high uranium density and thermal conductivity. However, its irradiation performance—particularly gas bubble swelling and defect dynamics—remains poorly characterized. Using in-situ transmission electron microscopy (TEM), we irradiated UC with 300 keV Xe⁺ and 1 MeV Kr²⁺ ions at temperatures up to 900 °C to quantify swelling behavior and dislocation loop evolution. The swelling remained below 0.6 % across all temperatures, suggesting the dimensional stability of UC under irradiation at these temperatures. Dislocation loops grew faster in UC than in UO₂ or UN, correlating with its lower homologous temperature. Notably, nanograin structures emerged in thin regions of the lamellar, mirroring phenomena previously observed in UO₂ and ZrC. These results address critical knowledge gaps in the radiation tolerance of UC and provide insight into its suitability for advanced reactor systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156082"},"PeriodicalIF":3.2,"publicationDate":"2025-07-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144860401","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Fabrication and characterization of high-performance Li2TiO3-Li4SiO4 biphasic ceramic tritium breeding cubic unit via digital light processing","authors":"Liang Cai ,&nbsp;Shurui Shang ,&nbsp;Haifeng Xue ,&nbsp;Xin Hu ,&nbsp;Qiang Qi ,&nbsp;Guangna Luo ,&nbsp;Yingchun Zhang","doi":"10.1016/j.jnucmat.2025.156066","DOIUrl":"10.1016/j.jnucmat.2025.156066","url":null,"abstract":"<div><div>Li₂TiO₃-Li₄SiO₄ biphasic ceramic has attracted increasing attention because of its unique advantages for tritium breeding in fusion reaction. To overcome the critical limitations of conventional Li₂TiO₃-Li₄SiO₄ pebbles, particularly their inadequate mechanical strength and low packing density, this work developed high-performance Li₂TiO₃-Li₄SiO₄ cubic units via digital light processing (DLP). The purge gas flow behavior in stacked cubic unit structure was studied by Computational Fluid Dynamics (CFD) method to evaluate its structural characteristics. The photosensitivity and rheological properties of precursor suspension were studied to verify the printing feasibility. Solid loading and sintering process were optimized to yield the defect-free Li₂TiO₃-Li₄SiO₄ cubic unit. The results show that the purge gas had periodic velocity distribution and extremely low pressure drop in the stacked cubic unit structure. Li₂TiO₃-Li₂SiO₃-Li₂CO₃ ceramic suspension possessed suitable photosensitivity and thixotropy, and its optimal solid loading was 40 vol%. After sintering at 1100 °C, the Li₂TiO₃-Li₄SiO₄ cubic unit with high relative density (92.8 %TD), high crush load (2000.3 N) and favorable open-pore structure can be prepared. Comparative analyses of Li-based cubic units indicate that the Li₂TiO₃-Li₄SiO₄ cubic unit not only addresses the limitations of single-phase variants but also demonstrates superior intrinsic characteristics, including excellent mechanical&amp; thermomechanical properties and distinct hydrogen isotope release behavior. This work proves the feasibility of Li₂TiO₃-Li₄SiO₄ biphasic strategy and potential of Li₂TiO₃-Li₄SiO₄ cubic units for the future tritium breeding design.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156066"},"PeriodicalIF":3.2,"publicationDate":"2025-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144750035","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"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 strategy for imparting radiation resistance to dilute alloys using synergistic solutes","authors":"Soumyajit Jana,&nbsp;Robert S. Averback,&nbsp;Pascal Bellon","doi":"10.1016/j.jnucmat.2025.156070","DOIUrl":"10.1016/j.jnucmat.2025.156070","url":null,"abstract":"<div><div>A novel approach for imparting radiation resistance to dilute alloys is proposed whereby two synergistic solute species are employed, a first one, solute B, that binds strongly to vacancies and a second one, solute C, that binds to solute B and is also a slow diffuser in solvent A. This combination results in B-C solute clusters that are immobile traps for vacancies. These traps promote point-defect recombination over irradiation doses far beyond that achievable in binary alloys, where solutes that strongly bind to vacancies are typically fast diffusers and thus quickly removed from grain interiors by radiation-induced segregation. A parametric study, performed using atomistic kinetic Monte Carlo simulations with realistic metallic solute properties in Cu, reveals that alloy stability under irradiation derives largely from the formation of mixed B-C solute clusters comprised of 10 or more atoms. The solute loss at sinks, moreover, is found to follow stretched exponentials, with the most promising alloys corresponding to values of the stretch exponent <span><math><mi>β</mi></math></span> approaching 0.5. The effects of irradiation dose rate and grain size are discussed using simple scaling relationships. Lastly, the approach is illustrated by identifying promising solute combinations in Cu, Ni and Al alloys.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156070"},"PeriodicalIF":3.2,"publicationDate":"2025-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144750371","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Implications of point defect accumulation on UO2 thermal conductivity and fission gas release under accelerated fuel irradiation","authors":"Mutaz Alshannaq ,&nbsp;Charlie Owen ,&nbsp;Joshua Ferrigno ,&nbsp;Md Minaruzzaman ,&nbsp;Hany S. Abdel-Khalik ,&nbsp;Anant Raj ,&nbsp;Tsvetoslav R. Pavlov ,&nbsp;Marat Khafizov","doi":"10.1016/j.jnucmat.2025.156057","DOIUrl":"10.1016/j.jnucmat.2025.156057","url":null,"abstract":"<div><div>Evaluation of thermal properties is a crucial factor for nuclear fuel performance. During reactor operation, the accumulation of fission products and irradiation-induced lattice defects are responsible for degradation in thermal conductivity. This affects fuel temperature and fission gas release (FGR) among other multiphysics processes important for economics and safety analysis. However, the point defects (PD) induced thermal conductivity reduction are not treated mechanistically in the current fuel performance codes (FPC). In this study, we analyze the implications of PD accumulation described using a rate theory (RT) model on the lattice thermal conductivity of UO₂ by adopting Lucuta thermal conductivity correlation (LC). We demonstrate that fission rate-dependent point defect concentrations have the largest impact on in-pile thermal conductivity in the periphery of light water reactor fuel below a temperature threshold governed by the migration barrier of defects. The reduction of thermal conductivity in the low -temperature rim region acts as additional thermal resistance and leads to a temperature notably larger than suggested by LC specifically at low burnups. These effects are anticipated to have notable impacts on fuel during the accelerated conditions. The impact of incorporating a point defect-informed approach to thermal conductivity is assessed through a detailed analysis of fission gas release behavior and fuel microstructure evolution. Finally, a fission rate-dependent correction to the Lucuta correlation is proposed as part of this analysis. The modified Lucuta correlation demonstrates higher FGR compared to the original correlation, although the accelerated irradiation process leads to a reduction in overall FGR. Once additional multiphysics mechanisms tightly coupled to temperature profile are introduced it becomes harder to deconvolve the impact of point defects.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156057"},"PeriodicalIF":3.2,"publicationDate":"2025-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144826911","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"International round robin onion irradiation of alloy T91 and comparison with neutron irradiation","authors":"G.S. Was ,&nbsp;C. Cabet ,&nbsp;C. Kaden ,&nbsp;M.H. Mayoral ,&nbsp;C. Pareige ,&nbsp;D. Bhattacharyya ,&nbsp;C. David ,&nbsp;C. Hardie ,&nbsp;D. Terentyev ,&nbsp;W. Weber ,&nbsp;T. Wei ,&nbsp;F. Naab ,&nbsp;V. Pauly ,&nbsp;I. Swainson ,&nbsp;M.S. Veshchunov","doi":"10.1016/j.jnucmat.2025.156065","DOIUrl":"10.1016/j.jnucmat.2025.156065","url":null,"abstract":"<div><div>An international collaboration was established as a Coordinated Research Project (CRP) under the IAEA and entitled Accelerator Simulation and Theoretical Modelling of Radiation Effects-II (SMoRE-II). It was created to determine, by way of a Round Robin process, the degree to which ion irradiations produced the same irradiated microstructure when conducted in different labs on the same alloy and provided with the same irradiation protocol. The Round Robin consisted of 13 participating organizations from 9 IAEA member states with ion irradiations conducted at all CRP partner sites on samples of a single alloy (T91) from a single billet with the same thermal-mechanical history, and with a specific protocol for conducting the irradiations. Of the 14 parameters specified for the ion irradiations, only 1 of 12 facilities was able to follow the protocol exactly. Major differences included vacuum pressure, temperature measurement and control, beam mode (raster-scanning vs. steady beam), and dosimetry. The microstructure features characterized were the sizes and number densities of cavities, dislocation loops, precipitates, and the radiation induced segregation. While loop size and number density appeared to correlate with carbon content, no such correlation was identified for cavities. The divergence from the irradiation protocol undoubtedly affected the irradiated microstructure with carbon contamination occurring in most cases. The cavity, dislocation loop and precipitate microstructures all fell within the range of that in the literature. Additionally, a T91 sample that was irradiated to 47 dpa at 376°C in the BOR-60 reactor was selected for comparison of the microstructure to those in the Round Robin study.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156065"},"PeriodicalIF":3.2,"publicationDate":"2025-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144773139","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Densification of Ba2NaIO6 ceramic wasteform for enhanced iodine immobilization","authors":"Yi Zhao ,&nbsp;Qu Ai ,&nbsp;Shi-Kuan Sun ,&nbsp;Sheng-Heng Tan ,&nbsp;Xiao Liang ,&nbsp;Wei-Ming Guo ,&nbsp;Hua-tay Lin","doi":"10.1016/j.jnucmat.2025.156062","DOIUrl":"10.1016/j.jnucmat.2025.156062","url":null,"abstract":"<div><div>Radioactive iodine-129 (¹²⁹I) presents a major challenge for nuclear waste management due to its long half-life and high environmental mobility. The design of iodine-bearing wasteforms is expected to balance the iodine loading with the chemical durability. Additionally, the volatile nature of iodine during high-temperature processing necessitates careful selection of wasteform type and accurate control of heat-treatment condition. This study introduces periodate-based Ba₂NaIO₆ double perovskites as the ceramic wasteform for immobilization of ¹²⁹I. Ba₂NaIO₆ powder was initially synthesized by solid state reaction at 650 °C, followed by densification using spark plasma sintering (SPS) at 900–1000 °C under vacuum. It was found that optimal densification was achieved at 950 °C, with a relative density of 97.69 %. Phase-assemblage and microstructural analyses confirmed the thermal stability of Ba₂NaIO₆ phase, which maintained <em>Fm-3</em> <em>m</em> double perovskites structure and minimal lattice distortion post-sintering. EDS mapping and composition analysis demonstrated the uniform distribution and effective incorporation of iodine in the matrix. XPS analysis revealed that iodine remained primarily in the +7 oxidation state, indicative of the chemical integrity during sintering. Leaching rate tests further displayed the excellent chemical stability, where the normalized iodine release rate of 1.65 ± 0.16 × 10^–⁴ g/(m²·d) after 7 days was determined. The sintering condition, iodine incorporation capacity, and dissolution rate were systematically compared with previously reported iodine-containing wasteforms.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156062"},"PeriodicalIF":3.2,"publicationDate":"2025-07-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144887116","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Microstructure evolution and deuterium permeation in RAFM steel under pulsed thermal loads","authors":"Zongming Shao ,&nbsp;Hao Yang ,&nbsp;Siwei Zhang","doi":"10.1016/j.jnucmat.2025.156067","DOIUrl":"10.1016/j.jnucmat.2025.156067","url":null,"abstract":"<div><div>The transport of hydrogen isotopes in reduced activation ferritic-martensitic (RAFM) steels under transient heat loads is crucial for their application as plasma facing materials (PFMs) in DEMO reactors. However, existing research predominantly focuses on hydrogen isotope retention within these materials, while permeation behaviour through RAFMs under transient heat loads remains poorly understood. In this study, China low activation martensitic (CLAM) steel as one of RAFMs was exposed in the QSPA-T facility to investigate the deuterium permeation behaviour under varying pulse numbers and heat load energy densities. Post-experiment characterization was performed using scanning electron microscopy (SEM), energy-dispersive spectroscopy (EDS), and electron backscatter diffraction (EBSD) to analyse the microstructural evolution and elemental distribution. And the deuterium permeation behaviour of the samples was determined by gas-driven permeation (GDP) experiments. The results show that the deuterium permeability of the sample increased by approximately 1∼2 orders of magnitude after heat loads, and the permeability increased with the number of the pulses, but showed significant change with the increase in energy density of the heat load. Combining the microstructural analysis of the samples after heat exposure, the observed phenomenon can be attributed to two main factors: an increase in surface roughness and the changes in microstructures of the resolidified layer.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"616 ","pages":"Article 156067"},"PeriodicalIF":3.2,"publicationDate":"2025-07-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144750033","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}