Nuclear Engineering and Design最新文献

{"title":"Experimental investigation of gas-phase migration behaviors in a 1 × 2 rod bundle under two-phase equilibrium and non-equilibrium flow","authors":"Hao Xie,&nbsp;Wenhai Qu,&nbsp;Jinbiao Xiong","doi":"10.1016/j.nucengdes.2026.114806","DOIUrl":"10.1016/j.nucengdes.2026.114806","url":null,"abstract":"<div><div>Void fraction gradient was assumed as the mechanism of void drift between subchannels. However, classical void drift models based on this assumption predict poor results under bubbly flow and cap-bubbly flow against experimental data. In this work, void fraction and bubble diameter in an enlarged 1 × 2 rod bundle was experimentally investigated by wire mesh sensors (WMS) system under equilibrium and non-equilibrium gas flow rates. The total 168 cases cover bubbly flow, cap-bubbly flow and churn flow defined with bubble shape and volume-base probability density function (VPDF). In general, small bubbles gather near channel walls, while large bubbles concentrate in sub-channel centers. For bubbly flow and cap-bubbly flow, it is difficult for bubbles passing through gap between subchannels. Thus, classical void drift models deviate from experimental results in bubbly flow and cap-bubbly flow. However, when bubble diameter is larger than 0.75 time of pitch of rod bundle, void drift improves obviously because bubbles are large enough to disturb the corresponding subchannel. Under bubbly flow and cap-bubbly flow, large bubbles coalesce in each subchannel. With gas superficial velocity increasing, void fraction and bubble diameter of large bubbles increase, while VPDF of small bubbles decreases. With liquid superficial velocity increasing, void fraction and bubble diameter of large bubbles decrease, while VPDF of small bubbles increases. For churn flow, strong void fraction migration between sub-channels is mainly caused by large bubbles with diameter larger than pitch of 32 mm. Thus, equilibrium distribution of void fraction can be fully developed within distance of 1830 mm under equilibrium and non-equilibrium inlet conditions. This is the reason that classical void drift models predict well against experimental data in churn flow. Large bubbles break down in churn flow. With gas superficial velocity increasing, void fraction and bubble diameter of large bubbles increase, while VPDF of small bubbles decreases. With liquid superficial velocity increasing, void fraction and bubble diameter of large bubbles decrease, while VPDF of small bubbles increases. This work helps to under mechanism of gas migration between sub-channels.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114806"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190735","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 state-of-the-art review of R&D for the super critical water-cooled reactor technology. Part I economics, thermalhydraulics, safety and licensing","authors":"Armando Nava Dominguez ,&nbsp;Chukwudi Azih ,&nbsp;Alberto D'Ansi Mendoza España ,&nbsp;Hussam Zahlan ,&nbsp;Guido Mazzini ,&nbsp;Alis Musa-Ruscak ,&nbsp;Sara Kassem ,&nbsp;Andrea Pucciarelli ,&nbsp;Walter Ambrosini ,&nbsp;Fabian Wiltschko ,&nbsp;Ivan Otic ,&nbsp;Tamás Varju ,&nbsp;Attila Kiss ,&nbsp;Pan Wu ,&nbsp;Elena Poplavskaia","doi":"10.1016/j.nucengdes.2026.114767","DOIUrl":"10.1016/j.nucengdes.2026.114767","url":null,"abstract":"<div><div>This study presents a summary of the most relevant research and development (R&amp;D) carried out to support the development of the only Generation IV water-cooled reactor endorsed by the Generation IV International Forum (GIF). The coolant of the proposed reactor is operated at supercritical water conditions, allowing for an increase in thermodynamic efficiency of the plant and the production of higher-grade process heat. Several collaborations have been established to support the development of this technology under the GIF umbrella, as well as through other international avenues. Therefore, the development is bolstered by a collective effort between numerous R&amp;D institutions across Asia, Europe, and North America. Globally, the R&amp;D programs have been methodologically executed in phases, namely: fundamental R&amp;D, validation and verification of assumptions used in R&amp;D analyses, and pre-conceptualization of supercritical water-cooled reactors (SCWRs).</div><div>This article summarizes the recent R&amp;D work performed to support the development of the SCWR technology on thermalhydraulics and safety, economics and licensing. Furthermore, R&amp;D highlights, identified knowledge gaps, conclusions, and recommendations are presented.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114767"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036056","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":"Theoretical model of fluid-structure interaction and prediction of fluidelastic instability for wire-wrapped fuel rods in axial flow","authors":"Shusheng Dai ,&nbsp;Xiaochang Li ,&nbsp;Yu Zhang ,&nbsp;Ruifeng Tian ,&nbsp;Jiming Wen ,&nbsp;Sichao Tan","doi":"10.1016/j.nucengdes.2026.114786","DOIUrl":"10.1016/j.nucengdes.2026.114786","url":null,"abstract":"<div><div>Wire-wrapped fuel rod bundles are commonly employed in Generation IV fast reactors. However, the swirl and perturbations induced by the helical wires can significantly enhance fluid-structure interaction and increase the risk of instability, and traditional models still struggle to capture the response characteristics. This study establishes a high-order theoretical model of fluid-structure interaction for wire-wrapped fuel rods in axial flow, grounded in the fundamental principles of Newtonian mechanics. Building upon the dynamic framework of a bare rod, the model incorporates the additional effects of wire-wrap mass, stiffness, and drag, enabling a concise yet systematic representation of wire-wrap effects. A dimensionless parameter system is established for nondimensionalization, and the Galerkin method is applied for discretization. The resulting system matrix is solved computationally to obtain the first wet-mode natural frequency and the critical instability velocity. The predicted results show good agreement with numerical simulations, with relative frequency errors below 8%, validating the model's reliability. Parameter sensitivity analysis is further performed to elucidate the effects of wire-wrap diameter and pitch on the critical instability velocity, and its coupled regulatory mechanisms on system stability under different rod diameters and lengths. The results indicate that increasing the wire-wrap diameter intensifies flow disturbances and reduces the critical instability velocity; a critical range of wire-wrap pitch exists that leads to the lowest system stability, displaying non-monotonic characteristics; and variations in rod parameters influence the wire-wrap effect on system stability, with smaller diameters or longer rods promoting instability, diameter reduction enhancing coupling, and length increase weakening it.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114786"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036024","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":"Effect of FDS uncertainty in fire simulations of nuclear power plants under different ventilation conditions","authors":"María González-Alvear ,&nbsp;Mariano Lázaro ,&nbsp;Daniel Alvear ,&nbsp;Eugenia Morgado ,&nbsp;Miguel Ángel Jiménez ,&nbsp;David Lázaro","doi":"10.1016/j.nucengdes.2025.114733","DOIUrl":"10.1016/j.nucengdes.2025.114733","url":null,"abstract":"<div><div>Fire Dynamics Simulator (FDS) is a well-known fire computer model, which has been widely applied for different scenarios. In particular, several standards and guidelines support its use in fire safety engineering approaches in nuclear plants. Although the uncertainty of the FDS model has been analysed and collected in literature, the influence of each input parameter has not yet been fully addressed.</div><div>Some of these previous contributions were based on the Benchmark Exercise No. 3 of the International Collaborative Fire Model Project (NUREG 6905). Moreover, the Best Practice Guidelines of NEA/CSNI/R (2014)11 were used as the reference to analyse the influence of the boundary conditions on simulation results. This work aims to study the impact of selected key fire dynamics parameters on the simulations of that scenario, updating previous findings.</div><div>Since this fire scenario involves three horizontal cable trays and one vertical cable tray, it is of special interest for nuclear power plants. Moreover, it is relevant to analyse the influence of input parameters on cable ignition. A sensitivity analysis was conducted, to evaluate the most important parameters for the selected scenario, focussing on ventilation and the thermal properties of the cables such as conductivity, specific heat, density, emissivity.</div><div>The results show the influence of each parameter in the surface temperature and heat flux in the different cable trays. Consequently, this enables the authors to formulate some recommendations for defining fire scenarios when applying fire safety engineering principles in nuclear power plants.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114733"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036030","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":"Risk-informed structural integrity assessment of the HTR-PM reactor pressure vessel","authors":"Bowen Li,&nbsp;Haitao Wang,&nbsp;Yanhua Zheng,&nbsp;Zhengming Zhang,&nbsp;Heng Peng","doi":"10.1016/j.nucengdes.2026.114764","DOIUrl":"10.1016/j.nucengdes.2026.114764","url":null,"abstract":"<div><div>The high-temperature gas-cooled reactor pebble-bed module (HTR-PM) in China represents the world's first demonstration power plant of its kind. As one of the Generation IV reactors, it is designed with inherent safety characteristics and offers potential for optimization in the structural integrity management of its reactor pressure vessel (RPV). Key advantages of the HTR-PM vessel, such as lower fast neutron flux than conventional pressurized water reactors, significantly slow thermal transients and reversed thermal gradients, collectively contribute to reducing both the risk and consequence of its failure. Using probabilistic fracture mechanics (PFM), a quantitative model has been developed to assess flaw initiation, crack propagation, and fracture of the RPV under typical HTR-PM transient and test conditions. This model enables assessment of the RPV failure probability, thereby establishing a basis for risk-informed safety evaluation and for optimization of both manufacturing and in-service inspection (ISI) requirements. Simulation results show that the RPV failure probability of HTR-PM is extremely low in the conservative case of no ISIs during its entire service life. Furthermore, flaws in circumferential weld regions contribute only minimally to the overall failure probability, suggesting potential for further extending inspection intervals or even exempting these areas from periodic inspection.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114764"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981783","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":"Gas bubble detection and segmentation using a machine learning approach leveraging semi-supervised training","authors":"J. Schäfer ,&nbsp;S. Taş ,&nbsp;U. Hampel","doi":"10.1016/j.nucengdes.2026.114763","DOIUrl":"10.1016/j.nucengdes.2026.114763","url":null,"abstract":"<div><div>This work presents a novel training approach for machine learning based instance segmentation, without the need for manual annotated datasets. With applications in experimental investigations in bubbly flows in reactor safety research. This semi-supervised process consists of two different neural networks, a conditional generative adversarial network used for data generation and a U-net style convolutional neural network for instance segmentation. We validated the approach using a fully automated experimental setup creating layered bubble curtains, which enables an evaluation of the bubble size distribution with and without overlapping bubbles. The final neural network is able to measure the average bubble size as well as to recreate the bubble size distribution accurately.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114763"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981787","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":"Integrated latent risk of radiological damage under conflict situation – Part II: Application and results","authors":"J.C. De la Rosa Blul ,&nbsp;M.Á. Hernández-Ceballos ,&nbsp;O. Parera Villacampa ,&nbsp;G. Magrotti ,&nbsp;A. Guglielmelli ,&nbsp;C. Fazio","doi":"10.1016/j.nucengdes.2025.114716","DOIUrl":"10.1016/j.nucengdes.2025.114716","url":null,"abstract":"<div><div>In situations where radiological risk does not originate from a single ongoing accident but from a widespread degradation of the operational context conditions affecting several nuclear installations, traditional event-oriented emergency preparedness may become insufficient. Under such conflict-related circumstances, protective planning requires a broader framework capable of capturing the aggregated and latent risk arising from multiple facilities and locations.</div><div>This paper presents the Diagnosis And Prognosis of Hazards in Nuclear Emergencies (DAPHNE) methodology (developed by the European Commission's Joint Research Centre), designed to support European Commission services and EU Member States in nuclear emergency preparedness and response. The methodology integrates MAAP, HYSPLIT, and the JRODOS decision-support system—together with dedicated in-house release models where appropriate—to provide a coherent assessment of both onsite accident progression and offsite radiological consequences.</div><div>Building on the conceptual framework introduced in Part I, this paper applies DAPHNE to a conflict-scenario case study to quantify integrated latent radiological risk and illustrate how the methodology enhances the identification of geographical areas where risk accumulates. By providing a harmonised numerical workflow for scenario selection, source-term calculation, dispersion modelling, and risk mapping, DAPHNE improves upon other methodologies and offers a scientifically grounded basis for designing conflict-related Emergency Planning Zones that complement existing site-specific arrangements.</div><div>Through its capacity to integrate multiple risk sources and their aggregated contribution to radiological exposure, DAPHNE strengthens preparedness planning under complex conflict scenarios and supports more informed decision-making across affected territories.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114716"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190813","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":"CIGMA experiments on integral phenomena related to thermal hydraulics in a reactor containment vessel and building during a severe accident","authors":"Satoshi Abe,&nbsp;Ari Hamdani,&nbsp;Shu Soma,&nbsp;Ryosuke Hangai,&nbsp;Masashi Ohmori,&nbsp;Akihiko Ohwada,&nbsp;Toshihito Ohmiya,&nbsp;Yasuteru Sibamoto","doi":"10.1016/j.nucengdes.2026.114787","DOIUrl":"10.1016/j.nucengdes.2026.114787","url":null,"abstract":"<div><div>The Fukushima Daiichi accident underscored the urgent need to understand complex thermal-hydraulic phenomena governing containment integrity and gas mixture distribution during a severe accident. In response, the Japan Atomic Energy Agency (JAEA) established the CIGMA (Containment InteGral Measurement Apparatus) facility, a flagship large-scale installation capable of high-temperature, high-pressure experiments with a steam-air‑helium gas mixture. This paper presents key findings from a comprehensive experimental campaign with CIGMA. The JT-SJ series demonstrated the effectiveness of external surface cooling in suppressing top head flange overheating. The CC-SP series revealed spray-induced mixing mechanisms that rapidly homogenize flammable stratifications. The CC-PL series identified condensation processes of the gas mixture that are decisive for containment cooling strategies. Finally, the CC-SJ series provided insights into inter-compartment gas transport relevant to the multi-stage explosions in Unit 3 of Fukushima Daiichi. These results establish a high-fidelity experimental database, offering benchmarks for CFD validation and advancing the development of robust hydrogen mitigation and accident management strategies worldwide.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114787"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036057","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 nitrogen dissolution characteristics in the nitrogen pressurization system of high flux reactors","authors":"Xinhe Qu,&nbsp;Yi Huang,&nbsp;Wei Peng,&nbsp;Heng Xie","doi":"10.1016/j.nucengdes.2026.114814","DOIUrl":"10.1016/j.nucengdes.2026.114814","url":null,"abstract":"<div><div>Many research reactors use a nitrogen pressurization system, where nitrogen is in direct contact with the primary coolant in the volume compensator, and the dissolved nitrogen molecules and the water molecules undergo radiolytic reactions such as ionization and excitation under the action of high-energy rays. The radiolytic products may affect the analysis and control of the primary coolant chemistry. Therefore, it is necessary to determine the solubility of nitrogen in the primary circuit. This study employed Aspen Plus to investigate and analyze nitrogen dissolution behavior in the primary circuit coolant of high flux reactors. First, the accuracy of the calculation model was validated by calculating the saturated solubility of nitrogen. Furthermore, steady-state simulations were performed for the primary circuit of a high flux reactor including a jet degassing process. Moreover, the key factors affecting the amount of nitrogen dissolved in the primary circuit, including the fluctuating flow rate in the surge line, the operating pressure of the degasser, the mass flow in the purification and degasification circuit (PDC), and the amount of nitrogen replenishment, were analyzed. The results indicate that nitrogen ingress from the volume compensator into the reactor pressure vessel and primary circuit occurs predominantly through convective flow, with diffusion playing a negligible role. The fluctuating flow rate in the surge line and the flow rate in the PDC have a significant influence on the amount of nitrogen dissolved in the primary circuit. Under typical analysis conditions, with a degasser pressure of 0.26 bar, a PDC flow rate at 1% of the primary circuit flow rate, and a fluctuating flow rate in the surge line of 0.1 kg/s, the nitrogen concentration in the primary circuit reaches 4.10 ppm—significantly below the saturated nitrogen solubility level. The results of this study have significant importance for the design of coolant pH control and the safety analysis of the high-flux reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114814"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190255","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":"Comparative analysis of machine learning methods for correcting uranium loading measurement errors in pebble-bed HTGR spherical fuel elements","authors":"Yanlong Wen,&nbsp;Hongjian Zhang,&nbsp;Haiyan Xiao,&nbsp;Qing Zhu,&nbsp;Liguo Zhang,&nbsp;Hong Li","doi":"10.1016/j.nucengdes.2026.114807","DOIUrl":"10.1016/j.nucengdes.2026.114807","url":null,"abstract":"<div><div>Accurate determination of uranium loading in spherical fuel elements of high-temperature gas-cooled reactors (HTGRs) is essential for nuclear material accounting and reactor safety. Conventional γ-ray spectrometry based on a single characteristic peak (185.7 keV) suffers from significant variability due to differences in fuel sphere dimensions and TRISO particle distributions, yielding a coefficient of variation (CV) of around 1%.</div><div>In this study, a machine learning–based error correction framework was developed to enhance UO₂ loading measurements. Three representative algorithms—Random Forest (RF), Support Vector Regression (SVR), and Multilayer Perceptron (MLP)—were trained on Geant4-simulated γ-ray data, incorporating both raw counts and composite features derived from six characteristic γ lines. Results show that all models reduced the CV to 0.3–0.4%, with the MLP achieving the best performance (CV ≤ 0.35%).</div><div>Further comparative analysis indicated that the three algorithms emphasize different mathematical solution spaces: RF captures discrete threshold effects, SVR provides smooth nonlinear fitting via kernel methods, and MLP flexibly spans a richer function space to approximate the mixed discrete–continuous nature of γ-ray attenuation. The integration of these perspectives demonstrates not only the potential of machine learning for nuclear material measurement but also the complementarity of different models in balancing accuracy, robustness, and interpretability.</div><div>This work provides a practical framework for improving non-destructive UO₂ loading measurements in HTGR fuel elements, supporting more reliable nuclear material accounting and contributing to advanced reactor fuel management.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114807"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190734","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}