Journal of Nuclear Materials最新文献

{"title":"Microstructural evolution of neutron irradiated ultrafine-grained austenitic stainless steel","authors":"Frederic Habiyaremye ,&nbsp;Solène Rouland ,&nbsp;Bertrand Radiguet ,&nbsp;Fabien Cuvilly ,&nbsp;Benjamin Klaes ,&nbsp;Benoit Tanguy ,&nbsp;Joël Malaplate ,&nbsp;Christophe Domain ,&nbsp;Diogo Goncalves ,&nbsp;Marina M. Abramova ,&nbsp;Nariman A. Enikeev ,&nbsp;Xavier Sauvage ,&nbsp;Auriane Etienne","doi":"10.1016/j.jnucmat.2025.155710","DOIUrl":"10.1016/j.jnucmat.2025.155710","url":null,"abstract":"<div><div>Austenitic stainless steels utilized in-core components of pressurized water reactors are prone to radiation-induced segregation, which leads to the degradation of microstructure and mechanical properties. To improve irradiation resistance, one possible solution is to increase the number density of point defect sinks, such as grain boundaries. For this purpose, ultrafine-grained or nanostructured microstructures are recommended due to their high density of grain boundaries. This paper investigates the microstructural changes in ultrafine-grained 316 austenitic stainless steel exposed to neutron radiation up to 3.9 dpa in irradiation conditions representative of light water reactors. The microstructure at different length scales was analyzed using electron backscattered diffraction, transmission electron microscopy, and atom probe tomography before and after neutron irradiation. The study compares its findings with those of existing literature on coarse-grained austenitic stainless steels to evaluate the benefit of ultrafine-grained 316 austenitic stainless steels regarding irradiation ageing in representative conditions of light water reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155710"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143512334","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"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":"Modeling and analysis for the anisotropic irradiation swelling of porous SiC/SiC composites","authors":"Luning Chen,&nbsp;Jing Zhang,&nbsp;Shurong Ding","doi":"10.1016/j.jnucmat.2025.155711","DOIUrl":"10.1016/j.jnucmat.2025.155711","url":null,"abstract":"<div><div>SiC/SiC composites are one of the promising engineering materials for nuclear applications. Anisotropic swelling deformations were observed in these materials during irradiation, and the underlying mechanisms should be deeply understood. In this study, a numerical simulation method is developed to predict the irradiation-induced deformations of the as-fabricated SiC/SiC composites. An emphasis is given to the generation of an RVE (Representative Volume Element) model with a pre-existing pore and the assumed residual stress field. Besides, the thermo-mechanical constitutive relations and stress update algorithms for the solid skeleton of porous SiC/SiC composites are developed with their irradiation effects considered comprehensively. Based on the homogenization theory, the calculation models to obtain the macroscopic swelling strains of porous SiC/SiC composites are developed. The good agreements between the predictions and the post-irradiation data of anisotropic swelling validate the effectiveness of the developed models and simulation methods. Research findings indicate that the irradiation creep deformations due to the existing residual stresses and high transient creep rate coefficients lead to the through-thickness size shrinkage of the pre-existing pores, which possibly becomes the dominant mechanism of the negative linear swelling of the SiC/SiC sample during the initial irradiation stage. The effects of the initial residual stress fields and the elastic constitutive relations on the anisotropic irradiation swelling behaviors are investigated. This study lays a foundation for the advanced manufacture of the SiC/SiC composites and the based multi-layer cladding tubes.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155711"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143512330","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":"Structure of the fuel-cladding chemical interaction (FCCI) layer of a high burnup Zr-1Nb nuclear fuel cladding","authors":"Wulin Song ,&nbsp;Xue Han ,&nbsp;Huanlin Cheng,&nbsp;Qi Tang,&nbsp;Huacai Wang,&nbsp;Songtao Ji","doi":"10.1016/j.jnucmat.2025.155699","DOIUrl":"10.1016/j.jnucmat.2025.155699","url":null,"abstract":"<div><div>A comprehensive characterization of the Fuel Cladding Chemical Interaction (FCCI) layer in a Zr-1Nb alloy with a burnup of 41GWd·tU⁻¹ has been performed utilizing primary techniques including Optical Microscopy (OM), Transmission Electron Microscopy (TEM), and Transmission Kikuchi Diffraction (TKD). The results indicate that the FCCI layer characterized in this study is mainly composed of tetragonal zirconia on both the cladding and fuel sides, with monoclinic zirconia in between. Additionally, the strong fiber texture in monoclinic and tetragonal zirconia aligns well with that in the water-side oxide film. The orientation relationships between α-Zr, monoclinic zirconia and tetragonal zirconia are (<span><math><mover><mn>1</mn><mo>¯</mo></mover></math></span>011) _α-Zr || (010) _m-ZrO2 and (10<span><math><mover><mn>1</mn><mo>¯</mo></mover></math></span>)_m-ZrO2 || (100) _t-ZrO2. It appears that the dominant force for texture development in the FCCI formed on this alloy is the α-Zr to m-ZrO₂ and m-ZrO₂ to t-ZrO₂ transformation stress which is independent with metal substrate orientation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155699"},"PeriodicalIF":2.8,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143512335","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":"Development of a dynamic-mesh porosity transport model for multi-dimensional fuel performance codes","authors":"Edoardo Luciano Brunetto ,&nbsp;Carlo Fiorina ,&nbsp;Andreas Pautz ,&nbsp;Sander van Til ,&nbsp;Fitriana Nindiyasari ,&nbsp;Alexander Fedorov ,&nbsp;Alessandro Scolaro","doi":"10.1016/j.jnucmat.2025.155717","DOIUrl":"10.1016/j.jnucmat.2025.155717","url":null,"abstract":"<div><div>The porosity redistribution within nuclear fuel pellets exposed to high power ratings plays a critical role in the thermo-mechanical behavior of fast reactor fuel. Traditional fuel performance codes predict porosity migration through advection-dominated transport equations often assuming a fixed geometry, and limiting their accuracy in asymmetric conditions. A novel dynamic-mesh porosity migration model has been developed to address these limitations. For verification and demonstration purposes, the model has been implemented in OFFBEAT, a multidimensional OpenFOAM-based fuel performance code. The solver dynamically adjusts the fuel pellet geometry to model the evolution of the central hole caused by pore migration. Mesh quality is preserved throughout the simulation by means dynamic-mesh algorithms involving the resolution of a mesh-motion equation to diffuse the displacement imposed at the mesh boundaries to all the domain points. The methodology incorporates modifications to the traditional porosity transport equation, correcting the advective fluxes in the governing equations to account for mesh points movement. A simple mechanistic model to determine the hole expansion velocity as a function of the local porosity, pore velocity and inner fuel radius is proposed. The model's parameters are calibrated using open literature experimental data, demonstrating the solver capability to predict central void diameters within acceptable discrepancy. The dynamic-mesh solver shows good accuracy in predicting off-centered hole formations and aligns well with post-irradiation examination data. This new approach preserves the foundational principles of existing porosity migration models while offering enhanced flexibility and accuracy in asymmetric heat transfer scenarios.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155717"},"PeriodicalIF":2.8,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143535253","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"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":"Microstructure, oxidation kinetics and hydrogen absorption of Cr-coated Zr-Sn-Nb alloy cladding tubes after single-sided oxidation at 1000–1200 °C followed by fast reflood","authors":"Weiwei Xiao ,&nbsp;Sheng Xu ,&nbsp;Xiao Hu ,&nbsp;Jinghao Huang ,&nbsp;Shihong Liu ,&nbsp;Shuliang Zou","doi":"10.1016/j.jnucmat.2025.155718","DOIUrl":"10.1016/j.jnucmat.2025.155718","url":null,"abstract":"<div><div>Reflood of nuclear fuel assemblies is the top priority accident management strategy for nuclear power plants in the event of a loss of coolant accident, during which the cladding tubes inevitably undergo reflood oxidation. This study aims to investigate the single-sided reflood oxidation behavior of Cr-coated Zr-Sn-Nb alloy cladding tubes at 1000 °C-1200 °C. High-temperature steam oxidation and in-situ quenching were employed to simulate the reflood oxidation process of nuclear fuel assembly cladding tubes in the early stages of severe accidents. The microstructure, cross-sectional layer thickness evolution, oxidation kinetics, and hydrogen absorption of Cr-coated Zr-Sn-Nb alloy cladding tubes during single-sided reflood oxidation process were investigated. The results showed that after single-sided reflood oxidation, microcracks appeared on the surface of the cladding tubes. As the oxidation temperature increases and the oxidation time prolongs, the surface oxidation products gradually evolve from porous flocculent structures to strip-shaped or elliptical bubble structures and worm aggregated structures. A multi-layer layered structure of Cr₂O₃ layer/Cr coating/Cr-Zr diffusion layer/α-Zr(O) was formed on the cross-section of the cladding tube after single-sided reflood oxidation. The thickness of the Cr₂O₃ layer and residual Cr coating does not increase or decrease monotonically with the extension of oxidation time after reflood oxidation at 1200 °C. The kinetics of single-sided reflood oxidation follows a parabolic law, and the oxidation constant increases by about an order of magnitude as the oxidation temperature increases by 100 °C. As the oxidation temperature increases and oxidation time prolongs, the hydrogen absorption of the cladding tube gradually increases. After single-sided reflood oxidation, the hydrides in the Zr-Sn-Nb alloy cladding tube are mainly δ-ZrH_1.5.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155718"},"PeriodicalIF":2.8,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143535341","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":"Fuel performance simulations of TRISO particle geometries derived from XCT","authors":"M. Poschmann,&nbsp;A. Prudil,&nbsp;R. Osmond","doi":"10.1016/j.jnucmat.2025.155714","DOIUrl":"10.1016/j.jnucmat.2025.155714","url":null,"abstract":"<div><div>The current AGR TRISO fuel specification effectively assumes that the layer thickness variations within a particle do not significantly affect particle performance. However, the limits of this assumption and their relevance for commercial TRISO production have not been established. In this work, a method was developed to generate 3D geometries of TRISO particles, including the spatial variation in layer thickness, from X-ray computed tomography for use in fuel performance modelling. Simulated irradiation of a demonstration particle found SiC hoop stress values peaking at 315 MPa in tension, significantly in excess of those from previous modelling studies with similar particle aspect ratios. Simulations with representative 2D axisymmetric geometries based on the demonstration particle predicted significantly lower stresses for the same simulated irradiation. 2D radial segments extracted with an arbitrarily oriented polar axis under-predicted the maximum SiC hoop stress by 315-400 MPa, while those extracted with the polar axis passing through the point of maximum SiC hoop stress in the 3D model under-predicted the maximum SiC hoop stress by 165-275 MPa. The 2D model produced using existing methods for generating a 2D flat-spot particle under-predicted the maximum SiC hoop stress by 215 MPa. These findings suggest that existing models may underestimate the stress caused by the asphericity of certain TRISO particle morphologies, and that the current AGR specification may not capture all of the geometric factors that contribute to particle failure probability.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155714"},"PeriodicalIF":2.8,"publicationDate":"2025-02-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143511580","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":"The effect of precipitates and alloying elements on γ-Fe (111) surface dissolution corrosion in liquid lead-bismuth eutectic by first-principles study","authors":"Yufei Li ,&nbsp;Runyu Zhou ,&nbsp;Tao Gao ,&nbsp;Changan Chen","doi":"10.1016/j.jnucmat.2025.155716","DOIUrl":"10.1016/j.jnucmat.2025.155716","url":null,"abstract":"<div><div>The M₂₃C₆ precipitates play an important role in the corrosion behavior of austenitic stainless steel. Here, by establishing the connection between the surface and the precipitated phase, this work assesses the surface dissolution corrosion by applying the electrode potential. Firstly, the influence of different carbides precipitates on surface dissolution corrosion is compared, which shows that Cr₂₂FeC₆ has the highest electrode potential (+1.68 V), accelerating surface dissolution corrosion. Then, the effect of solute atoms (Pb/Bi/O) on surface dissolution corrosion is studied. It is found that Pb/Bi will promote the dissolution of surface Fe atoms. However, O will strengthen the corrosion resistance of the surface. Simultaneously, the O inhibits the hybridization of the 3d orbital of Fe and the 6p orbital of Bi/Pb, mitigating the corrosion of Pb/Bi on the surface. Lastly, the effects of common alloying elements (Al, Si, and Ni) in austenitic steel on the corrosion of the surface are also investigated to improve surface corrosion resistance. This research attempts to provide a more comprehensive knowledge of the corrosion of iron substrates in ADSs, improving the safety of nuclear energy systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155716"},"PeriodicalIF":2.8,"publicationDate":"2025-02-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143528685","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 self-ion (Fe+) irradiation of ODS-FeCrAl alloy fuel cladding materials with different Cr contents: The early stages of Cr-rich α’ phase precipitation","authors":"Hoang Le ,&nbsp;Yann de Carlan ,&nbsp;David T. Hoelzer ,&nbsp;Kan Sakamoto ,&nbsp;Per O.Å. Persson ,&nbsp;Jonathan A. Hinks ,&nbsp;Konstantina Lambrinou","doi":"10.1016/j.jnucmat.2025.155706","DOIUrl":"10.1016/j.jnucmat.2025.155706","url":null,"abstract":"<div><div>Oxide-dispersion-strengthened FeCrAl (ODS-FeCrAl) alloys are candidate accident-tolerant fuel cladding materials for light water reactors because they demonstrate satisfactory resistance to materials degradation effects such as high-temperature oxidation, radiation-induced swelling, and creep. Their perspective deployment to market is challenged, however, by their inherent susceptibility to irradiation embrittlement caused by the precipitation of the brittle Cr-rich α’ phase at relatively low temperatures (≤475 °C). This work used <em>in situ</em> self-ion irradiation (150 keV Fe⁺) in a transmission electron microscope to elucidate the early stages of Cr-rich α’ phase precipitation in three candidate ODS-FeCrAl alloy fuel cladding materials with different Cr contents (10, 12, and 20 wt.%) and microstructures. The early stages of the process resulting in the precipitation of the Cr-rich α’ phase in these three ODS-FeCrAl alloys under Fe⁺ irradiation were investigated at room temperature and 300 °C up to total fluences of 1.7 × 10¹⁵ ions·cm^-2 (2 dpa) and 3.4 × 10¹⁵ ions·cm^-2 (4 dpa), using three damage dose rates (5 × 10^–5, 3.3 × 10^–4, and 2 × 10^–3 dpa·s^-1). Post-irradiation examination via scanning transmission electron microscopy, energy-dispersive X-ray spectroscopy and electron energy loss spectroscopy suggested that the precipitation of the Cr-rich α’ phase might be promoted by the phase separation of the alloy matrix into Cr-rich and Fe-rich regions. Interestingly, oxygen impurities segregated preferentially in the Cr-rich regions, possibly promoting the radiation-assisted formation of the Cr-rich α’ phase. α’ phase precipitation was more pronounced at room temperature when compared to 300 °C, and it was clearly promoted by the progressive increase in the Cr content of the ODS-FeCrAl alloy.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155706"},"PeriodicalIF":2.8,"publicationDate":"2025-02-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143535245","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"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":"Experimental study on the oxidation behavior of nuclear graphite IG-110","authors":"Bin Lin ,&nbsp;Liang Chen ,&nbsp;Guisen Liu ,&nbsp;Yao Shen","doi":"10.1016/j.jnucmat.2025.155712","DOIUrl":"10.1016/j.jnucmat.2025.155712","url":null,"abstract":"<div><div>This study presents an experimental investigation on the oxidation behavior of nuclear graphite IG-110. A gas chromatography-based oxidation test platform was developed to monitor the oxidation processes of IG-110 graphite in different atmospheres. The oxidation tests focused on the presence of H₂O and O₂, with temperatures up to 750 °C and various partial pressures. The graphite consumption was analyzed, and a Boltzmann function was proposed to describe the relationship between graphite consumption and oxygen partial pressure. A graphite consumption prediction model was developed based on this function, which can be utilized to calculate the graphite consumption of CO/CO₂ reaction products in High-Temperature Gas-cooled Reactors (HTGRs).</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155712"},"PeriodicalIF":2.8,"publicationDate":"2025-02-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143511579","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":"LBE erosion-corrosion behaviors of gelcasted high-speed rotating Ti3AlC2 impeller","authors":"Zhanchong Zhao ,&nbsp;Shibo Shi ,&nbsp;Xinxin Gao ,&nbsp;Qingsheng Wang ,&nbsp;Youyuan Zhang ,&nbsp;Yuying Wen ,&nbsp;Yang Ling ,&nbsp;Xian Zeng ,&nbsp;Mei Ma","doi":"10.1016/j.jnucmat.2025.155708","DOIUrl":"10.1016/j.jnucmat.2025.155708","url":null,"abstract":"<div><div>The MAX phase stands out as one of the highly prospective candidate materials for the impeller of the lead-bismuth fast reactor nuclear main pump. In order to determine the compatibility between the high-speed rotating Ti₃AlC₂ ceramic impeller and the liquid lead-bismuth eutectic (LBE) alloy, the corrosion behavior was comprehensively investigated within the LBE environment. In this study, the erosion-corrosion behavior of a high-speed rotating Ti₃AlC₂ ceramic impeller in LBE at 550 °C for 600 h was systematically investigated. The research results indicate that the Ti₃AlC₂ impeller maintained its structural integrity without macroscopic fractures after dynamic corrosion. Weight evaluation shows that during the corrosion test, the degradation of its physical or chemical properties was negligible. Surface analysis revealed the formation of a corrosion layer primarily composed of TiC and amorphous carbon, with PbO adhering to the surface as a protective barrier, effectively mitigating Ti and Al losses. Comparative analysis confirmed the superior adhesiveness of PbO over TiO₂ and Al₂O₃. The impeller maintained its mechanical performance, experiencing only a minor weight gain of 0.02 wt.% during the test. Vibration analysis confirmed operational stability, with a maximum stress of 10.76 MPa and a rotational frequency of 16.7 Hz, well below the first-order resonance frequency of 4997.7 Hz. This study furnishes crucial insights into the corrosion characteristics of the MAX phase and presents significant data. It is anticipated to provide a valuable reference for the initial application of MAX-phase ceramic impellers in advanced nuclear systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155708"},"PeriodicalIF":2.8,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143487305","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}