Nuclear Engineering and Design最新文献

{"title":"A generalized coarse mesh finite difference acceleration for 3D method of characteristics in hexagonal fast reactor simulations","authors":"Xiaoxu Geng ,&nbsp;Yun Hu ,&nbsp;Jiayi Yu ,&nbsp;Chong Wang ,&nbsp;Haodong Shan ,&nbsp;Zhaoshun Wang","doi":"10.1016/j.nucengdes.2025.114429","DOIUrl":"10.1016/j.nucengdes.2025.114429","url":null,"abstract":"<div><div>Numerical simulation has become essential in nuclear reactor design and safety verification due to the high cost and complexity of physical experiments. The Method of Characteristics (MOC) provides high-fidelity neutron transport solutions with inherent parallelism and geometric flexibility. However, conventional MOC implementations face challenges in memory usage, computational efficiency, and convergence speed, particularly for full-core simulations of fast reactors with complex hexagonal assemblies. To address these challenges, this work extends the self-developed 3D neutron transport solver ANT-MOC by implementing a hexagonally track generation module that reduces track counts and memory demands while improving accuracy. The developed generalized coarse mesh finite difference (GCMFD) method, compared with CMFD, naturally supports unstructured hexagonal/pentagonal meshes and enables efficient mesh indexing, adjacency management, and equivalent width calculations in such geometries. In addition, an energy group coarsening acceleration framework is introduced to alleviate the computational burden caused by the fine energy discretization typical of fast reactors. These features make the method particularly suitable for fast reactor simulations with complex geometries and wide energy spectra. Validation on the China Experimental Fast Reactor (CEFR) full-core benchmark shows the framework achieves an <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>e</mi><mi>f</mi><mi>f</mi></mrow></msub></math></span> error of 18.4<!--> <!-->pcm and fission rate and scalar flux errors below 0.3%. The iteration count decreases from 411 to 144, significantly enhancing convergence efficiency. These results demonstrate a high-precision and efficient neutron transport simulation capability for fast reactor cores, with strong potential for engineering applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114429"},"PeriodicalIF":2.1,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145046817","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":"State prediction and analysis of 3D upper plenum of lead–bismuth fast reactor based on model order reduction under transient accidents","authors":"Wenshun Duan ,&nbsp;Carolina Introini ,&nbsp;Antonio Cammi ,&nbsp;Kefan Zhang ,&nbsp;Sifan Dong ,&nbsp;Hongli Chen","doi":"10.1016/j.nucengdes.2025.114447","DOIUrl":"10.1016/j.nucengdes.2025.114447","url":null,"abstract":"<div><div>Accurate prediction of three-dimensional (3D) thermal–hydraulic parameter evolution during transients in lead–bismuth fast reactors is important for safety. Although high-fidelity computational fluid dynamic (CFD) models are accurate, they are computationally expensive for real-time use. Model order reduction (MOR) techniques can alleviate this cost while retaining accuracy. In this work, the upper plenum of the lead–bismuth fast reactor NCLFR-Oil is taken as the object of study. Using the proper orthogonal decomposition (POD)-based MOR method and artificial neural networks (ANN), two different 3D transient analysis frameworks are proposed for different data scenarios. 1) A time-series hybrid model (THM) framework designed for time multiple-query tasks, which enables rapid prediction of future three-dimensional physical fields through nonlinear temporal extrapolation of reduced-order modal coefficients. 2) A hybrid data assimilation (HDA) framework aimed at situations with limited sensor data, where the full 3D field distribution is reconstructed using only sparse temperature measurement points by integrating real-time sensor observations with the MOR. The frameworks enhance computational efficiency significantly, with maximum errors around 0.05. Speed-up ratios of 940 and 713 are achieved for THM and HDA frameworks, respectively. Using only three noisy temperature sensors, the HDA framework accurately reconstructs pressure, temperature, and velocity fields, demonstrating robustness and practical applicability. Sensitivity analyses further confirm reliability under varying sensor numbers and noise levels. This work provides an effective tool for real-time monitoring and safety evaluation under accident conditions, offering high practical value.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114447"},"PeriodicalIF":2.1,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145027522","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":"Evaluation of the efficiency of heat supply from nuclear power plants taking into account the heat storage properties of heating networks","authors":"Michael Garievsky,&nbsp;Elena Burdenkova","doi":"10.1016/j.nucengdes.2025.114463","DOIUrl":"10.1016/j.nucengdes.2025.114463","url":null,"abstract":"<div><div>The increasing share of nuclear generation in energy systems necessitates enhanced efficiency and flexibility of nuclear power plants (NPPs), within the bounds of engineering constraints. This article explores ways for increasing NPP efficiency through cogeneration and by using the heat storage properties of district heating networks. It is shown that with the development of centralized heat supply based on nuclear power plants, it is possible not only to reduce fossil fuel consumption and greenhouse gas emissions, but also enhance the operational flexibility of nuclear power plants without altering the reactor power output. Based on thermodynamic calculations of thermal circuits of NPP turbines and economic analysis, a comprehensive assessment of the efficiency of heat supply from NPPs using turbines with uncontrolled steam extraction was performed. Special attention is paid to the thermal storage capacity of heating networks, which allows for the accumulation of thermal energy and redistribution of the heat load over a 24-hour period. Calculations show that utilizing the heat storage properties of heating networks for flexible operation of NPPs increases profits by 5.3 %. For various methods of laying heating networks, the maximum distances from the nuclear power plant to the heat energy consumers at which capital investments pay off have been determined.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114463"},"PeriodicalIF":2.1,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145046816","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 framework for state assessment and remaining useful life prediction of control rod drive mechanism roller","authors":"Mengqi Huang ,&nbsp;Zhengyu Du ,&nbsp;Ruibo Lu ,&nbsp;Xiaoji Wang ,&nbsp;Changhong Peng","doi":"10.1016/j.nucengdes.2025.114460","DOIUrl":"10.1016/j.nucengdes.2025.114460","url":null,"abstract":"<div><div>The control rod drive mechanism (CRDM) roller is susceptible to degradation, such as wear and fatigue, under environmental and operational stresses, including temperature, humidity, friction, and impact. To enable timely operating state diagnosis, predict future degradation trends, and support operation control and maintenance decisions, this study develops a framework for CRDM roller state assessment and remaining useful life (RUL) prediction. In the state assessment stage, residual distribution analysis combined with adaptive neighbourhood radius clustering is employed to evaluate roller states under limited input data. In the RUL prediction stage, a particle filter-based degradation model is constructed, with parameters estimated via maximum likelihood and multi-objective optimization, and further corrected using Bayesian theory and backward smoothing to enhance prediction accuracy. Validation on bearing datasets achieved a degradation onset estimation deviation of less than three time steps and a RUL prediction error of 8.19%. At the same time, the application to the CRDM roller yielded an error of 8.96%. These results confirm the framework’s capability for accurate state diagnosis and precise RUL prediction using real-time monitoring data.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114460"},"PeriodicalIF":2.1,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145027616","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":"Finding leaking fuel in the core based on 134Cs and 137Cs activities during spiking events","authors":"I.A. Evdokimov ,&nbsp;D.V. Dmitriev ,&nbsp;A.G. Khromov ,&nbsp;E.Y. Afanasieva ,&nbsp;P.M. Kalinichev ,&nbsp;A.A. Sorokin ,&nbsp;I.O. Goryushin ,&nbsp;A.Y. Burtsev ,&nbsp;S.P. Zolotarev ,&nbsp;S.V. Babkin ,&nbsp;T.Y. Kvichanskaya ,&nbsp;V.V. Atrazhev","doi":"10.1016/j.nucengdes.2025.114439","DOIUrl":"10.1016/j.nucengdes.2025.114439","url":null,"abstract":"<div><div>At present, the ratio of ¹³⁴Cs and ¹³⁷Cs activities during spiking events is widely considered to be the best indicator of fuel burnup in leaking fuel rods. Evaluations are performed by comparison of the ratio between ¹³⁴Cs and ¹³⁷Cs activities in primary coolant with a reference function of fuel burnup. However, there is no universal correlation between the ¹³⁴Cs/¹³⁷Cs activity ratio and fuel burnup. Since ¹³⁴Cs is produced through neutron capture, the ¹³⁴Cs/¹³⁷Cs activity ratio in fuel depends on neutron energy spectrum. The spectrum is sensitive to fuel enrichment and burnup as well as to fission, moderation and absorption characteristics of the surrounding environment. For this reason, the ¹³⁴Cs/¹³⁷Cs activity ratio as a function of burnup differs for different fuel types and changes from cycle to cycle every time when the fuel loading pattern in the core is varied. Using a unified correlation in practice introduces significant errors in the evaluation of leaking fuel burnup. A new approach is developed for the identification of leaking FAs in the core. In this approach, ¹³⁴Cs and ¹³⁷Cs inventory is calculated with low computational costs for each fuel rod in reactor. A software application has been developed to automatically identify fuel rods in the core that provide the closest agreement with the ¹³⁴Cs/¹³⁷Cs activity ratio measured during the spiking event. The developed approach was validated on NPP data for 14 fuel cycles, each containing a single leaking FA. None of the 14 leaking FAs matched the burnup range predicted by the standard technique, with difference ranging from ∼ 10 to 35<!--> <!-->MWd/kgU. In contrast, the new approach accurately identified all 14 leaking FAs and prioritized them for leakage testing during reactor outage.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114439"},"PeriodicalIF":2.1,"publicationDate":"2025-09-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020687","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":"Cesium removal from aqueous solutions using mineral adsorbents: Mechanisms, kinetics, and thermodynamics","authors":"Shizong Wang ,&nbsp;Jianlong Wang ,&nbsp;Xuan Guo","doi":"10.1016/j.nucengdes.2025.114461","DOIUrl":"10.1016/j.nucengdes.2025.114461","url":null,"abstract":"<div><div>This study evaluated cesium (Cs) removal by six minerals, with removal efficiencies at pH 7 of 100 % (zeolite), 85.4 % (montmorillonite), 83.8 % (Na-bentonite), 56.2 % (bentonite), 29.5 % (K-feldspar), and 24.3 % (phlogopite). Sips model-derived capacities reached 15.1  mg/g (zeolite), 11.3 mg/g (montmorillonite), 9.67 mg/g (Na-bentonite), 9.02 mg/g (bentonite), 5.07 mg/g (K-feldspar), and 4.70 mg/g (phlogopite). Thermodynamic analysis confirmed spontaneous, endothermic adsorption. Mechanisms analysis revealed that Cs removal occurred mainly via ion exchange or surface coordination, varying by mineral. Zeolite immobilized Cs in its porous structure; montmorillonite and bentonite used interlayer exchange and hydroxyl coordination; phlogopite formed Al-O-Cs complexes via hydroxyl-fluoride substitution; Na-bentonite enabled Cs-Na exchange with octahedral [Al(OH)₆]^3– stabilization; K-feldspar achieved Cs-O-Al surface bonding. Under 100  kGy ⁶⁰Co irradiation, phlogopite, Na-bentonite, and K-feldspar maintained stable Cs adsorption, while zeolite, montmorillonite, and bentonite showed efficiency reductions of ∼ 8–27 %. Among the tested materials, zeolite, montmorillonite, and Na-bentonite are recommended for their high Cs affinity and radiation durability. These findings highlight the importance of balancing adsorption capacity and radiation resistance in selecting optimal minerals for radioactive Cs remediation and long-term environmental protection.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114461"},"PeriodicalIF":2.1,"publicationDate":"2025-09-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020686","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":"3D modeling of passive safety condensers with CATHARE3 for a high-power pressurized water reactor","authors":"Michel Belliard,&nbsp;Lucie Groussy","doi":"10.1016/j.nucengdes.2025.114391","DOIUrl":"10.1016/j.nucengdes.2025.114391","url":null,"abstract":"<div><div>In the wake of the Fukushima’s nuclear accident, passive safety systems have gained in appeal to improve a nuclear power plant’s resistance to accidents without requiring external power supplies. SAfety COndensers (SACO) are a promising example of these new systems. These are secondary heat exchangers, immersed in a tertiary pool, attached to the steam piping of Steam Generators (SG). In the event of a failure in the normal water supply to the SG, they take the place of emergency pumps, which require a power supply. Their purpose is to extract residual power from the core by condensation of the secondary steam produced at the SG, and return it in liquid form. Then, the secondary liquid inventory is preserved, preventing the SG from drying out.</div><div>For several years now, CEA, in collaboration with EDF, has been involved in modeling SACO of various designs (straight or “C”-shaped vertical tubes, in a calandria or a small or large pool, etc.) for new reactor concepts. In particular, the 3D modeling of immersed exchangers in a pool, using the CATHARE3 code, challenges the conventional 0D/1D modeling and shows the interest of 3D spatial discretization to better take into account the lateral feed of exchangers. As a result, several 3D SACO models, based on different experimental designs and adapted to the reactor power under consideration, are proposed. There are compared with each other on a typical secondary depressurization transient. Also, the CATHARE3 3D modeling is discussed on a typical station black-out transient for a given type of SACO design.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114391"},"PeriodicalIF":2.1,"publicationDate":"2025-09-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020683","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":"Thermal effects on gas migration in saturated bentonite under rigid boundary conditions","authors":"Sai Li ,&nbsp;Weimin Ye ,&nbsp;Qian Zhang ,&nbsp;Qiong Wang ,&nbsp;Yonggui Chen","doi":"10.1016/j.nucengdes.2025.114462","DOIUrl":"10.1016/j.nucengdes.2025.114462","url":null,"abstract":"<div><div>The influence of temperature on gas migration behavior constitutes a critical consideration for both the design and operational safety of deep geological repositories for high-level radioactive wastes (HLW). In this study, water injection and subsequent gas injection tests were performed on specimens with dry densities of 1.3, 1.5 and 1.7 Mg/m³ at temperatures 20, 40 and 60 °C. Meanwhile, the specimens that experienced related gas injection tests were subjected to mercury intrusion porosimetry (MIP) tests. The results indicate that the effects of temperature on the effective gas permeability are dependent on both the injection pressure and the initial dry density. Under low gas injection pressures, an increase in temperature leads to a rise in effective gas permeability, while under high gas injection pressures, the temperature impact on the effective gas permeability depends on dry density. For specimens with high dry densities, the effective gas permeability positively correlates with temperature, while for low dry densities, the value at 40 °C exceeds those at 20 and 60 °C. Additionally, the gas breakthrough pressure decreases with increasing temperature. Higher dry density specimens are more likely to experience capillary breakthrough, while interfacial breakthrough more commonly happens in lower dry density specimens. According to the microstructural observations from the MIP tests, increasing temperature reduces the specimen pore space due to the shrinkage of the bentonite matrix. These findings indicate that gas migration in saturated bentonite is governed by a competitive mechanism between pore structure and the state of pore water, while both of which are influenced by temperature.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114462"},"PeriodicalIF":2.1,"publicationDate":"2025-09-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020682","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":"Performance of dry electrostatic precipitator for the removal of CsI aerosol particles under simulated severe nuclear accident conditions","authors":"Satish Basnet ,&nbsp;Atte Jäntti ,&nbsp;Pasi Yli-Pirilä ,&nbsp;Miika Kortelainen ,&nbsp;Olli Sippula ,&nbsp;Anna Lähde","doi":"10.1016/j.nucengdes.2025.114440","DOIUrl":"10.1016/j.nucengdes.2025.114440","url":null,"abstract":"<div><div>The performance of electrostatic precipitators (ESPs) in reducing the emission of radioactive aerosols and hydrogen was investigated under varying experimental conditions to improve the safety features of nuclear power plants (NPPs). The effects of gaseous atmosphere, humidity, flow rate, and temperature on particle properties and filtration efficiency were evaluated using caesium iodide as a model aerosol. The results indicate that the ESP achieved a maximum particle mass filtration efficiency of over 90 % for the lab-scale ESP and more than 99.5 % for the industrial ESP. However, the particle number concentration varied with the industrial ESP, highlighting effective removal of larger particles that acted as a condensation sink, allowing higher concentrations of newly formed ultrafine particles to persist in the system. Hydrogen mitigation experiments revealed no measurable impact of ESPs on hydrogen concentrations under typical operating conditions, confirming their safety within current NPP protocols. The study highlights the crucial role of particle properties, carrier gas composition, and ESP design in determining filtration efficiency, emphasising the need for further research on reactor-specific conditions to optimise ESP performance and enhance source term reduction strategies.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114440"},"PeriodicalIF":2.1,"publicationDate":"2025-09-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020684","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":"An integral effect test and its code simulation on the complete loss of reactor coolant system flowrate for the SMART100 design","authors":"Jin-Hwa Yang,&nbsp;Byong-Guk Jeon,&nbsp;Hwang Bae,&nbsp;Hyun-Sik Park","doi":"10.1016/j.nucengdes.2025.114431","DOIUrl":"10.1016/j.nucengdes.2025.114431","url":null,"abstract":"<div><div>A complete loss of reactor coolant system flowrate (CLOF) scenario was successfully tested using the integral effect test facility of SMART-ITL. The steady-state conditions were well achieved to satisfy initial test conditions presented in the test requirement; its boundary conditions were accurately simulated, and the CLOF scenario was reproduced properly for the SMART100 design. The CLOF test was performed over a long period of 60,000 s to understand the long-term behavior of the passive residual heat removal system (PRHRS) of the SMART100 design. The test results from the SMART-ITL were analyzed using the MARS-KS code to assess its capability to simulate a CLOF scenario for the SMART100 design. As the passive safety systems will operate for no less than 72 h to fulfill their function, the precise prediction of thermal–hydraulic behaviors in reactor coolant system (RCS) and passive safety systems of the SMART100 in the long-term sense is very crucial for the safety analyses of the SMART100 design. The measured thermal–hydraulic data from the CLOF test using the SMART-ITL were properly simulated using the MARS-KS code, which showed its reasonable simulation capability for the CLOF scenario of the SMART100 design with PRHRS. The thermal–hydraulic phenomena depend more on flow resistance both in RCS and PRHRS loops and heat losses from the RCS, steam generator, PRHRS loop and emergency cooldown tank during the long-term simulation. It is considered that accurate simulation is possible with proper consideration on flow resistance in the loops and heat loss through the heat structure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114431"},"PeriodicalIF":2.1,"publicationDate":"2025-09-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145020685","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}