Progress in Nuclear Energy最新文献

{"title":"Research on the impact of important parameters of component cooling system on design","authors":"Weiguang Zhao ,&nbsp;Pei Yu ,&nbsp;Xiaobo Zeng ,&nbsp;Guangming Fan ,&nbsp;Zhaoming Meng ,&nbsp;Changqi Yan","doi":"10.1016/j.pnucene.2025.105779","DOIUrl":"10.1016/j.pnucene.2025.105779","url":null,"abstract":"<div><div>There is a significant gap between the component cooling system (CCS) of the HPR1000 nuclear island cold chain system and the actual operating parameters, resulting in insufficient utilization of the cooling source and poor economic performance. Due to the high complexity of CCS design and the difficulty of design evaluation, multi-objective optimization research is necessary to identify parameters that significantly impact the design results of the CCS. To this end, this paper constructs a mathematical model and evaluation program for the CCS, utilizing an improved genetic algorithm suitable for solving constrained complex multi-objective optimization problems. The design temperature of the CCS and several large design flow rates are selected as optimization variables, while weight, volume, investment cost, and energy consumption are set as optimization objectives. The aim is to use optimization design methods to determine the impact of important parameters of the CCS on its design. The results indicate that, under the premise of ensuring the safety of nuclear power plants, the optimization algorithm can effectively identify the optimal combinations of optimization variables, leading to relative optimal solutions for the four objectives: a maximum reduction of 18.1 % in weight, 21.3 % in volume, 11.5 % in investment cost, and 28.7 % in energy consumption. Furthermore, the study finds that the design temperature of the CCS is a crucial parameter affecting its design results, and that appropriately reducing the design temperature is a key direction for optimizing the CCS design.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":""},"PeriodicalIF":3.3,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791221","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental study on heat transfer characteristics of subcritical carbon dioxide in a vertical tube","authors":"Jinghui Wu,&nbsp;Haicai Lyu,&nbsp;Cailu Xu,&nbsp;Xianjun Chen,&nbsp;Han Wang","doi":"10.1016/j.pnucene.2025.105777","DOIUrl":"10.1016/j.pnucene.2025.105777","url":null,"abstract":"<div><div>Research is actively being conducted globally to investigate the supercritical carbon dioxide (S-CO2) Brayton cycle which is considered a highly promising energy conversion system. During startup, load change, or shutdown processes, the system experiences subcritical pressure conditions where the phase change of carbon dioxide (CO2) may occur. Understanding these heat transfer characteristics becomes particularly crucial as the system transitions through subcritical regions, where abrupt property variations and phase change effects could significantly impact operational stability and safety. To address these challenges, the heat transfer properties of subcritical CO2 in a vertically arranged tube with a heated length of 750 mm and a diameter of 4 mm were investigated in this work. The effects of heat flux, pressure, and mass flux on the heat transfer characteristics were analyzed. Results indicated that with increased heat flux, heat transfer deterioration (HTD) occurs, leading to a peaky wall temperature. Increasing pressure reduces the surface tension of CO2 leading to earlier HTD. Additionally, higher mass flux slightly improves heat transfer coefficient under saturated boiling. Four correlations were selected to estimate the accuracy of predicting the heat transfer coefficient by the test result. The Fang correlation (Fang, 2013) was shown to exhibit the most accurateness than other correlations considered with average deviation and standard deviation of 2.42 % and 43.98 %, respectively. Improved predictions in the heat transfer coefficient were achieved by fitting a new coefficient based on the Fang correlation The critical heat flux to trigger HTD was obtained, and an empirical correlation was proposed.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105777"},"PeriodicalIF":3.3,"publicationDate":"2025-04-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143785825","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":"Transient characteristics of supercritical carbon dioxide Brayton cycle system with or without control strategy under accident conditions","authors":"Ziwen Su ,&nbsp;Xiaoyu Wang ,&nbsp;Yang Ming ,&nbsp;Fulong Zhao ,&nbsp;Ersheng You ,&nbsp;Sichao Tan ,&nbsp;Ruifeng Tian","doi":"10.1016/j.pnucene.2025.105772","DOIUrl":"10.1016/j.pnucene.2025.105772","url":null,"abstract":"<div><div>Supercritical carbon dioxide (SCO₂) Brayton cycle reactor system is a popular type of small modular reactor system. It has many benefits, such as high thermal efficiency, system structure simply and compact. So it can be widely used in various scenarios. In this study, establish the SCO₂ Brayton cycle system model which have multiple control models based on Modelica language. Transient conditions with and without control strategies were simulated for reactivity introduction accidents and external load loss accidents, among which the control strategies were bypass control, rotor speed control and throttle control respectively. The response characteristics of system variables in different accidents without control or under different control strategies are compared and analyzed. Then, the control effect of different control strategies is judged by the change amplitude of system parameters. Results display that the rotor speed control and throttle control can reduce the variation amplitude of system parameters in both accidents, rotor speed control can reduce the variation amplitude of representative parameters by more than 50 %, the control effect is better. Bypass control can reduce the amplitude of changes of most system variable in reactivity introduction accidents, the variation amplitude of partial parameters decreased by 7.69 %–34.17 %. But it can not play the role of reducing the changes of system variable in external load loss accidents. Therefore, the control effect of bypass control should be judged according to the actual situation. The research results can a provide reference for the study about formulation and selection of control strategies for SCO₂ Brayton cycle systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105772"},"PeriodicalIF":3.3,"publicationDate":"2025-04-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143785826","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":"Prediction of core unmeasurable parameters during loss of coolant accident using deep neural network method","authors":"Milad Moradi,&nbsp;Mohsen Ghafari","doi":"10.1016/j.pnucene.2025.105760","DOIUrl":"10.1016/j.pnucene.2025.105760","url":null,"abstract":"<div><div>The parameters within the reactor core would be categorized into two different groups of measurable and unmeasurable parameters. The determination of unmeasurable parameters, such as void fraction and critical heat flux, plays a significant and fundamental role in predicting the occurrence of accidents and emergency situations within the reactor. The utilization of deep neural networks represents one of the methods for accurate and reliable estimation of these parameters. Such estimations facilitate the implementation of necessary measures to prevent accidents or mitigate their consequences. In this study, three deep neural network models namely LSTM, TFT, and NBEATS are employed for void fraction prediction within the reactor core after Loss of Coolant Accident (LOCA). The neural network training will be performed without covariates, using past covariates and using future covariates. The results reveal that the TFT neural network, trained with future covariates (e.g. pressure, temperature, water velocity and steam velocity) yields the lowest error.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":""},"PeriodicalIF":3.3,"publicationDate":"2025-04-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143776598","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 review on properties required for FB-CVD fabricated SiC for TRISO fuel production","authors":"Linyi Hu ,&nbsp;Joel Turner ,&nbsp;James Buckley ,&nbsp;Tim Abram","doi":"10.1016/j.pnucene.2025.105762","DOIUrl":"10.1016/j.pnucene.2025.105762","url":null,"abstract":"<div><div>In pursuit of achieving Net-Zero by 2050, there is a demand for a stable, sustainable source of green energy. The renewed focus on a ‘nuclear renaissance’ has highlighted the importance of demonstrating High Temperature Gas-cooled Reactor (HTGR) technology by the early 2030s. The performance of HTGRs hinges on the quality of their advanced fuel form known as TRi-structural ISOtropic (TRISO) coated fuel particles.</div><div>Although the fabrication of TRISO particles achieved success in the late 20th century, the lack of commercial HTGR systems limited the development of this advanced fuel form, particularly in the UK after the termination of the Dragon Project. The small number of active research facilities globally has resulted in a lack of fundamental mechanistic understanding of the fabrication process, contributing to uncertainties in scaling up production to an industrial level.</div><div>The commissioning of an active Fluidised Bed Chemical Vapour Deposition (FB-CVD) coater at the University of Manchester underscores the necessity for a preliminary literature study on fabrication parameters (deposition temperature, gas composition, flow rate, etc.) crucial for achieving high-quality TRISO products, while considering the economical utilisation of experimental uranic materials. This study is presented in two parts focusing on silicon carbide (SiC) and pyrolytic carbon (PyC) layers respectively.</div><div>In the present work, SiC properties (stoichiometry, microstructure, density) and process parameters (coating rate, efficiency) are scrutinised. Experimental findings from historical and recent FB-CVD coaters worldwide have been analysed to establish correlations between fabrication parameters, SiC properties, and process outcomes. Consensus exists regarding the influence of deposition temperature, though conflicting results are observed concerning the impact of H₂/MTS. Other crucial fabrication parameters such as input gas flow rates remain to be definitively determined.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105762"},"PeriodicalIF":3.3,"publicationDate":"2025-04-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143777502","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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":"Effect of Al2O3, RHF, and RHA on gamma shielding and mechanical properties of TeO2-Based glass using Phy-X/PSD","authors":"Floressy Juhim ,&nbsp;Fuei Pien Chee ,&nbsp;Asmahani Awang ,&nbsp;Saafie Salleh ,&nbsp;Muhammad Izzuddin Rumaling ,&nbsp;Amani Alalawi ,&nbsp;M.S. Al-Buriahi","doi":"10.1016/j.pnucene.2025.105767","DOIUrl":"10.1016/j.pnucene.2025.105767","url":null,"abstract":"<div><div>Radiation attenuation behaviour of TeO₂–ZnO–Bi₂O₃–Na₂O–Er₂O₃ with the addition of Al₂O₃ nanoparticles (NP), rice husk fibre (RHF), and rice husk ash (RHA) was investigated using Phy-X/PSD software. The mass attenuation coefficients of the chosen glasses were measured from 0.015 to 3 MeV. The results showed that increasing the RHF concentration of the glasses or decreasing the Al₂O₃ component of the glass system enhances the mass attenuation coefficient of the material. At low energy, there is a significant difference between the mass attenuation coefficients values of the samples with the lowest and highest Al₂O₃ content, where the difference is 0.05118 cm²/g at 0.015 MeV. TZNERHF had higher mass attenuation coefficients (0.05402 cm²/g), and TZNETiAl8 had the highest mass attenuation coefficients, with a value of 0.05357 cm²/g at 1.173 MeV. The results showed a decreasing trend in the half-value layer (HVL), tenth value layer (TVL), and mean free path (MFP) when the density increased from 5.12555 to 5.23882 g/cm³ and attenuation ability improved. Subsequently, the influence of Al₂O₃ on Zeff values became more considerable as Zeff values increased. The shielding qualities of their glass samples are also superior to commercial window glass and common shielding concrete. Thus, the glass samples can be used with excellent advantage as radiation shielding materials. The findings showed that adding RHF and lower Al₂O₃ enhanced the studied samples' radiation shielding properties. The results show that the TZNERHF glass sample has the highest mechanical properties. Thus, the TZNERHF sample is better than the other glass samples.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105767"},"PeriodicalIF":3.3,"publicationDate":"2025-04-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143777504","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":"Assessment of RELAP5/SCDAPSIM code against NACIE-UP facility tests","authors":"Daniel Dupleac,&nbsp;Roxana-Mihaela Nistor-Vlad","doi":"10.1016/j.pnucene.2025.105775","DOIUrl":"10.1016/j.pnucene.2025.105775","url":null,"abstract":"<div><div>During the 2017 campaign, ADP10 and ADP06 tests were performed in NAtural Circulation Experiment – Upgraded (NACIE-UP) facility built at ENEA Brasimone Research Centre to analyze the behavior of the 19-pins wire-spaced fuel pin bundle simulator (FPS) during the transition from forced to natural circulation flow. Although the total heating power is the same in both tests, in the ADP10 test the power was uniformly distributed among the 19 electrically heated pins, whereas in the ADP06 test, only the 7 central pins were switched-on. The present paper presents the application of RELAP5/SCDAPSIM/MOD4.1 (RS4.1) code for ADP10 and ADP06 tests subchannel analysis. RELAP/SCDAPSIM code is a system code initially designed for the thermal-hydraulic and safety analysis of Light Water Reactors (LWRs). Over the years, the code has been extensively improved by incorporating additional fluid properties used as coolants for advanced reactors, among which molten lead, lead-bismuth eutectic (LBE) and molten salts. Although the RS4.1 is a system code the present paper describes the code capabilities in predicting the subchannels thermal behavior under the ADP10 and ADP06 test conditions from the NACIE-UP facility. RS4.1 simulations results for LBE mass flow rate and temperature measured by thermocouples showed good agreement with experimental data which indicates that RS4.1 code can be successfully employed as subchannel code.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105775"},"PeriodicalIF":3.3,"publicationDate":"2025-04-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143777503","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 flow heat transfer characteristics and mechanism of lithium-lead and helium-xenon","authors":"Zhuoran Lei ,&nbsp;Ruibo Lu ,&nbsp;Xiaojian Ge ,&nbsp;Haoran Shen ,&nbsp;Fulong Zhao ,&nbsp;Sichao Tan ,&nbsp;Ruifeng Tian","doi":"10.1016/j.pnucene.2025.105774","DOIUrl":"10.1016/j.pnucene.2025.105774","url":null,"abstract":"<div><div>To find out the coupling heat transfer problem between liquid lithium-lead (Li-Pb) and helium-xenon (He-Xe) in a Printed Circuit Heat Exchanger (PCHE), a three-dimensional model of a double-straight-through PCHE heat exchange unit was established. This paper employs numerical simulation to investigate the coupled heat transfer characteristics as well as the mechanism between liquid lithium-lead and helium-xenon and to fit the flow heat transfer correlation of lithium-lead under specific working conditions. The results show that the coupled heat transfer of liquid lithium-lead is mainly dominated by thermal conduction, mainly influenced by the wall temperature in contact with the fluid and the heat flux density at the wall. Convective heat transfer is the main mechanism influencing helium-xenon's coupled heat transfer. Modifying the inlet velocity of helium-xenon has a more significant effect on the heat exchanger's performance compared to adjusting the inlet velocity of liquid lithium-lead. Regardless of which side's temperature is changed, the heat transfer amount will be greatly affected, and the heat exchanger's overall heat transfer coefficient is mainly determined by the helium-xenon.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105774"},"PeriodicalIF":3.3,"publicationDate":"2025-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143760302","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":"Non-condensable gas accumulation and distribution due to condensation in the CIGMA facility: Implications for Fukushima Daiichi Unit 3 (1F3)","authors":"Ari Hamdani ,&nbsp;Shu Soma ,&nbsp;Satoshi Abe ,&nbsp;Yasuteru Sibamoto","doi":"10.1016/j.pnucene.2025.105771","DOIUrl":"10.1016/j.pnucene.2025.105771","url":null,"abstract":"<div><div>This study was motivated by previous analysis by TEPSYS, which indicated that different temperatures in the 5th and 4th floors of the reactor building (R/B) of Fukushima Daiichi Unit 3 (1F3) could significantly influence the distribution of non-condensable gases during the severe accident in 2011. Understanding this effect is crucial for assessing the risks associated with non-condensable gas accumulation in R/B, particularly considering that the hydrogen explosion in 1F3 may not have been a single-stage event but could have involved a multi-stage explosion. To investigate the distribution and transport on non-condensable gas under varying cooling conditions, an experimental study was conducted using the CIGMA facility, a large-scale test vessel with segmented sections created by partition plates, designed to replicate the structure and conditions of the R/B. Steam and helium, used as non-condensable hydrogen substituted, were continuously injected at the top of the CIGMA vessel for 10,000 s to replicate the leakage of steam and hydrogen through the shield plug. Two cooling conditions were tested: Case 1 with a cooling temperature of 50 °C, and Case 2 with a cooling temperature of 90 °C. The experiments aimed to explore how varying cooling temperatures impact helium accumulation in different regions of the R/B structure. The study found that in both cases, the highest concentration of non-condensable gases may not always be near the injection point but rather in downstream regions. In these regions, steam and non-condensable gases move downward and continue to condense, significantly enhancing the concentration of helium. Specifically, in Case 1, after 10,000 s, the helium concentration reached 65 % in the middle region (representing the 4th floor of R/B) and 45 % in the top region (representing the 5th floor of R/B). Analysis using the Shapiro ternary diagram showed that, under these conditions, the gas mixture in the middle region posed a potential detonation risk. This study provides important insights into gas distribution within a nuclear reactor building during severe accident conditions, which are crucial for developing more effective safety measures and risk mitigation strategies in nuclear reactor designs.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105771"},"PeriodicalIF":3.3,"publicationDate":"2025-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143760301","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 load following control strategy for heat pipe-cooled microreactor coupled with closed Brayton cycle system","authors":"Yiran Qian,&nbsp;Tao Liu,&nbsp;Yan Zhang,&nbsp;Yuyang Leng,&nbsp;Tianrun Shi,&nbsp;Meihui Song,&nbsp;Weixiong Chen","doi":"10.1016/j.pnucene.2025.105768","DOIUrl":"10.1016/j.pnucene.2025.105768","url":null,"abstract":"<div><div>Heat pipe-cooled reactors (HPCRs), with their straightforward and compact design and inherent safety features, are considered a promising option for mobile nuclear power sources. However, the coupling between the HPCR and the energy conversion system complicates load tracking control. In this study, a comprehensive dynamic model of an HPCR coupled with a closed-air Brayton cycle (CABC) is developed. This model integrates a neutron kinetics point reactor model, a thermal resistance network model for the heat pipe, and a CABC model representing the power conversion system. To achieve effective load tracking, a control scheme for the Brayton cycle power is proposed, combining turbine bypass control with inventory control. The system's response to load variations, from 100 % full power (FP) to 50 % FP at a rate of 5 % FP/min, is analyzed, considering reactor power self-regulation. The results indicate that the reactor power exhibits an overshoot of −4.36 % and stabilizes after 2100s, while the fuel average temperature fluctuates within ±4.6 °C and stabilizes at 2240s. Building on the dynamic performance of the HPCR-CABC system, a reactor power control scheme is developed by adjusting control drums to introduce external reactivity. This scheme successfully controls reactor power at a steady state within 720s, without fluctuation. However, the fuel average temperature fluctuates by more than 10 °C, stabilizing after 9450s. To improve control effectiveness, a temperature correction channel is incorporated into the reactor power control system. As a result, the fuel average temperature fluctuates within ±0.3 °C and stabilizes at the design value of 699.4 °C within 860s. Compared to previous control schemes, the improved cascade reactor power control scheme results in a shorter stabilization time for the fuel average temperature. To assess the adaptability of the proposed HPCR-CABC coordination control strategy, simulations are conducted for load reductions from 100 % FP to 80 %, 60 %, and 40 % FP at a variation rate of 20 % FP/min. The results demonstrate that the proposed HPCR-CABC coordination control strategy exhibits strong load-following capability.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105768"},"PeriodicalIF":3.3,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143738250","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}