Nuclear Engineering and Design最新文献_第5页

Leakage mechanism and dynamic response characteristics of double-seal O-ring structures in nuclear power engineering 核电工程中双密封o形环结构泄漏机理及动态响应特性

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-17 DOI: 10.1016/j.nucengdes.2025.114239

Lei He, Guoliang Xu, Ming Li, Xiaoming Huang

{"title":"Leakage mechanism and dynamic response characteristics of double-seal O-ring structures in nuclear power engineering","authors":"Lei He, Guoliang Xu, Ming Li, Xiaoming Huang","doi":"10.1016/j.nucengdes.2025.114239","DOIUrl":"10.1016/j.nucengdes.2025.114239","url":null,"abstract":"<div><div>This paper presents a flow resistance model for O-ring seals and establishes a MATLAB/SIMULINK simulation solver for predicting the dynamic response of a double-seal structure. The calculated values of the flow resistance at varying compression rates and gas pressures were evaluated and compared to the experimental data to validate the utility of the flow resistance model. The simulation solver is then applied to predict the dynamic leakage behavior of the double-seal structure under applied constant pressure steps and sinusoidal excitations, revealing their response characteristics as exponential and sinusoidal, respectively. The analysis further demonstrates that the key parameters of the dynamic response—including the time constant of the exponential response and the amplitude and phase of the sinusoidal response—are functions of the product of the flow resistance and the gas capacitance. Approximate analytical solutions for the dynamic leakage behavior of the double-seal structure under the two excitations are obtained by constructing an equivalent gas circuit model and neglecting the effects of state parameters on flow resistance and gas capacitance. A quantitative analysis indicates that the deviation between the approximate and simulated solutions is within 10%. The findings presented in this paper can inform the development of strategies to enhance the sealing performance of double-seal structures.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114239"},"PeriodicalIF":1.9,"publicationDate":"2025-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144297606","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}

引用次数: 0

Modelling of bubble formation phenomenon at top-submerged nozzles with side hole submerged in a quiescent liquid 边孔浸没在静液中的顶部浸没喷嘴气泡形成现象的模拟

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-17 DOI: 10.1016/j.nucengdes.2025.114108

Soumya Sarkar , Nirvik Sen , K.K. Singh

{"title":"Modelling of bubble formation phenomenon at top-submerged nozzles with side hole submerged in a quiescent liquid","authors":"Soumya Sarkar , Nirvik Sen , K.K. Singh","doi":"10.1016/j.nucengdes.2025.114108","DOIUrl":"10.1016/j.nucengdes.2025.114108","url":null,"abstract":"<div><div>Efficiency of a gas-injection based passive liquid mixing device depends on size of gas bubbles produced as it affects bubble rise velocity and hence gas hold-up volume. Typically, such devices employ gas dispersion through a foot-piece having side holes immersed in a liquid. Despite of their relevance, such devices and phenomena relevant to such devices have not been explored sufficiently. A key relevant phenomenon is bubble formation at a side hole in a nozzle immersed in a quiescent liquid. Therefore, an experimental study is conducted to investigate this phenomenon. Liquids used in the study are relevant to radiochemical plants which need such passive mixing devices. An empirical correlation is obtained to estimate bubble size using the experimental data. Absolute average relative deviation between predicted and measured bubble diameter values is ∼ 9 %. To develop a generalised framework to predict bubble diameter, force balance based mathematical modelling is also attempted. Validation of the mathematical model with experimental data is carried out. Average absolute relative deviation for this validation is ∼ 16 %. A parametric analysis using the mathematical model is also conducted to examine how various independent variables influence bubble size. The impact of these independent variables on bubble size is assessed through sensitivity analysis.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114108"},"PeriodicalIF":1.9,"publicationDate":"2025-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144297608","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}

引用次数: 0

Criticality and sensitivity/uncertainty analysis of the DNRR with LEU fuel using MCNP6.3 and the latest data libraries 使用MCNP6.3和最新数据库进行低浓铀燃料DNRR的临界性和敏感性/不确定性分析

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-17 DOI: 10.1016/j.nucengdes.2025.114182

Duc-Tu Dau , Nhi-Dien Nguyen , Ton-Nghiem Huynh , Quang-Huy Pham , Kien-Cuong Nguyen , Thoi-Nam Chu , Van-Khanh Hoang , Giang T.T. Phan , Huu-Tiep Nguyen , Thanh Mai Vu , Hoai-Nam Tran

{"title":"Criticality and sensitivity/uncertainty analysis of the DNRR with LEU fuel using MCNP6.3 and the latest data libraries","authors":"Duc-Tu Dau , Nhi-Dien Nguyen , Ton-Nghiem Huynh , Quang-Huy Pham , Kien-Cuong Nguyen , Thoi-Nam Chu , Van-Khanh Hoang , Giang T.T. Phan , Huu-Tiep Nguyen , Thanh Mai Vu , Hoai-Nam Tran","doi":"10.1016/j.nucengdes.2025.114182","DOIUrl":"10.1016/j.nucengdes.2025.114182","url":null,"abstract":"<div><div>Criticality and sensitivity/uncertainty analysis was conducted for the Dalat Nuclear Research Reactor (DNRR) with 92 low enriched uranium (LEU) fuel bundles using the MCNP6.3 code and the latest nuclear data libraries such as ENDF/B-VIII.0, JENDL-5, JEFF-3.3 and CENDL-3.2. Criticality analysis was conducted for thirty critical conditions of the DNRR corresponding to different control rod positions established experimentally in comparison with the measurements and among the data libraries. A good agreement was found between the calculations and experiments with the discrepancy of the <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>e</mi><mi>f</mi><mi>f</mi></mrow></msub></math></span> less than 354 pcm. Whereas, the discrepancies among the data libraries are within 172 pcm. The uncertainties of the <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>e</mi><mi>f</mi><mi>f</mi></mrow></msub></math></span> caused by the data libraries of ENDF/B-VIII.0, JENDL-5, JEFF-3.3 are 415.7, 363.0 and 588.0 pcm, respectively. Based on the analysis, improvements in cross-section evaluations for H-1, U-235, Al-27, and Be-9, particularly regarding the capture and elastic scattering of H-1 and the fission of U-235, are recommended to enhance the reliability of core analysis for the DNRR. The evaluations have identified ENDF/B-VIII.0 and JENDL-5 libraries as the most reliable and should be prioritized for future core physics and safety analyses of the DNRR and similar reactor.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114182"},"PeriodicalIF":1.9,"publicationDate":"2025-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144297609","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}

引用次数: 0

Dynamic mechanical properties of bonded-orthogonal trapezoidal honeycomb aluminum under lateral loading 横向载荷作用下粘接正交梯形蜂窝铝的动态力学性能

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-17 DOI: 10.1016/j.nucengdes.2025.114246

Youdong Xing , Yukun An , Siyi Yang , Zhen Cui , John Zhai

{"title":"Dynamic mechanical properties of bonded-orthogonal trapezoidal honeycomb aluminum under lateral loading","authors":"Youdong Xing , Yukun An , Siyi Yang , Zhen Cui , John Zhai","doi":"10.1016/j.nucengdes.2025.114246","DOIUrl":"10.1016/j.nucengdes.2025.114246","url":null,"abstract":"<div><div>In order to ensure the safety of nuclear spent fuel transportation casks during transit, shock absorption devices are installed at both ends of the cask. These devices are filled with shock-absorbing materials designed to dissipate energy during drops. For nuclear equipment, ideal shock-absorbing materials exhibit bidirectional or triaxial performance. This study investigates the dynamic mechanical properties of bonded orthogonal trapezoidal honeycomb aluminum (BOTHA), a bidirectional structural material (Y and Z directions have identical cross-sectional shapes；This material boasts a straightforward manufacturing process, cost-effectiveness, and comparable load-bearing performance on both sides, thereby equipping it to better address complex operational conditions.). We subjected three distinct BOTHA materials, differing in cell thickness −to-cell size ratios (0.3/2.5, 0.4/2.5, and 0.5/2.5), various velocity loads (1.66 × 10<sup>−3</sup>/s, 47/s and 290/s) along the Y or Z axis (lateral direction) to determine their deformation and energy absorption characteristics. The energy absorption diagram, derived from experimental results, allows for quick analysis of the deformation and energy absorption of the impact limiter under various operational conditions. Material parameter values were extracted according to the dynamic constitutive law. Through simulation, we further explored the material’s deformation and stress curve characteristics, vividly illustrated by the deformation cloud map. Finally, we assessed the feasibility of using BOTHA as an impact limiter material for nuclear equipment by comparing its shock absorption performance with traditional wood-based filling materials.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114246"},"PeriodicalIF":1.9,"publicationDate":"2025-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144297607","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}

引用次数: 0

MELCOR model of iPWR and application to emergency planning zone study iPWR MELCOR模型及其在应急规划区研究中的应用

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-16 DOI: 10.1016/j.nucengdes.2025.114226

John Cui, Luke Lebel, Andrew Morreale, David Hummel

{"title":"MELCOR model of iPWR and application to emergency planning zone study","authors":"John Cui, Luke Lebel, Andrew Morreale, David Hummel","doi":"10.1016/j.nucengdes.2025.114226","DOIUrl":"10.1016/j.nucengdes.2025.114226","url":null,"abstract":"<div><div>This study presents the MELCOR modelling and analysis of a generic integral pressurized water reactor (iPWR) system with the goal of informing a broader study on emergency planning zone (EPZ) sizing. The iPWR design is one of the small modular reactor (SMR) technologies that have high technical readiness and are being considered for potential near term deployment. To understand the possible iPWR accident scenarios/behaviours important to support emergency preparedness, this study uses the severe accident analysis code MELCOR. A station blackout (SBO) accident was selected as the base case along with other postulated accident scenarios such as loss of steam generator feedwater, failure of the decay heat removal system, breaks at operating pool, and breaks in the containment vessel. The analyses investigate accident progression in the reactor vessel, containment, and operating pool/reactor building, as well as radionuclide releases for the assessed cases to provide insights on EPZs. The simulated cases in this study provide a spectrum of mechanistic source terms that can fit into the broader probabilistic framework of accident frequency that could inform decisions on EPZ sizing.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114226"},"PeriodicalIF":1.9,"publicationDate":"2025-06-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144291247","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}

引用次数: 0

The effects of strain amplitude and temperature on kinematic hardening parameters for low cycle fatigue of AISI316L stainless steel 应变幅值和温度对AISI316L不锈钢低周疲劳运动硬化参数的影响

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-16 DOI: 10.1016/j.nucengdes.2025.114221

Sushant Bhalchandra Pate , Gintautas Dundulis , Stephan Courtin , Jean-Christophe Le Roux

引用次数: 0

Design and analysis of self-actuated shutdown system in a small lead–bismuth cooled fast reactor 小型铅铋冷快堆自驱动停堆系统设计与分析

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-16 DOI: 10.1016/j.nucengdes.2025.114225

Xin Jin , Lipeng Wang , Hui Guo , Lixin Chen , Xinbiao Jiang , Hanyang Gu

{"title":"Design and analysis of self-actuated shutdown system in a small lead–bismuth cooled fast reactor","authors":"Xin Jin , Lipeng Wang , Hui Guo , Lixin Chen , Xinbiao Jiang , Hanyang Gu","doi":"10.1016/j.nucengdes.2025.114225","DOIUrl":"10.1016/j.nucengdes.2025.114225","url":null,"abstract":"<div><div>The safe operation of small lead-based fast reactors (LFRs) requires ensuring that the core temperature remains within a safe range. High temperatures will intensify the corrosive effect of lead–bismuth (LBE) on structural materials, seriously affecting the core’s lifespan. By burnable poisons (BPs) design, the risk of fuel and fuel rod cladding overheating under CRW accident in the LFR-180 core can be reduced. However, the core still faces the corrosion challenges induced by elevated LBE temperatures during and following the ULOF and ULOHS scenarios. In this research, the feasibility of the application of the self-actuated shutdown system (SASS) is explored, and the function of SASS is examined under anticipated transients without scram (ATWS). The calculation results demonstrate that the SASS can effectively mitigate the core equilibrium temperature following transients, thereby significantly reducing the risk of severe corrosion of structural materials. Designed in combination with SASS and BPs, the LFR-180BP-SASS improves the safety of the core during and following ULOF, ULOHS, and CRW events.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114225"},"PeriodicalIF":1.9,"publicationDate":"2025-06-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144297605","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}

引用次数: 0

A reverse scatter correction method for CT images of nuclear graphite components 核石墨组分CT图像的反向散射校正方法

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-16 DOI: 10.1016/j.nucengdes.2025.114236

Tianchen Zeng , Jintao Fu , Peng Cong , Ximing Liu , Guangduo Xu , Yuewen Sun

{"title":"A reverse scatter correction method for CT images of nuclear graphite components","authors":"Tianchen Zeng , Jintao Fu , Peng Cong , Ximing Liu , Guangduo Xu , Yuewen Sun","doi":"10.1016/j.nucengdes.2025.114236","DOIUrl":"10.1016/j.nucengdes.2025.114236","url":null,"abstract":"<div><div>The structural integrity of high-temperature gas-cooled reactors (HTGRs) is reliant on over 3000 carbon/graphite components, each of which necessitates computed tomography (CT) scanning for non-destructive testing prior to deployment, as stipulated by the reactor’s technical specifications. Despite the critical role of CT scans, they are frequently marred by significant scatter artifacts due to the detection of scattered photons, which compromises image uniformity and diminishes the system’s ability to detect defects. Our research presents a novel inverse scatter photon prediction model that addresses the shortcomings of traditional models by utilizing empirical data and the inherent properties of nuclear graphite/carbon components. This method begins by mapping and subtracting the transmitted photons from the projection data, followed by estimating the scattered photon distribution based on their low-frequency characteristics. The experimental results confirm that our method not only surpasses Monte-Carlo simulations and machine learning in reducing scatter artifacts but also meets the efficiency requirements for online detection. Our quantitative analyses indicate that our approach achieves the highest Defects Recognition Performance (DRP) value, which underscores a notable enhancement in the system’s ability to detect defects. We are confident that our findings will significantly enhance the defect detection capabilities of the CT scanning system, thereby contributing to the overall safety and reliability of HTGR operations.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114236"},"PeriodicalIF":1.9,"publicationDate":"2025-06-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144297610","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}

引用次数: 0

High-temperature oxidation failure in reactivity-initiated accidents: An evaluation of failure criteria based on oxygen concentration from the previous NSRR experiments 反应性引发事故中的高温氧化失效：基于先前NSRR实验中氧浓度的失效标准评估

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-16 DOI: 10.1016/j.nucengdes.2025.114222

Vu-Nhut Luu, Yoshinori Taniguchi, Yutaka Udagawa, Jinya Katsuyama

{"title":"High-temperature oxidation failure in reactivity-initiated accidents: An evaluation of failure criteria based on oxygen concentration from the previous NSRR experiments","authors":"Vu-Nhut Luu, Yoshinori Taniguchi, Yutaka Udagawa, Jinya Katsuyama","doi":"10.1016/j.nucengdes.2025.114222","DOIUrl":"10.1016/j.nucengdes.2025.114222","url":null,"abstract":"<div><div>For near-term application, coated-Zr alloy claddings show potential for enhancing safety by providing better oxidation resistance and minimizing hydrogen absorption under design-basis accidents (DBA). This benefit could extend the burnup and operational cycles of fuel rods. In assessing safety, reactivity-initiated accidents (RIA) are considered as one of the DBA conditions. The current safety criteria for high-temperature oxidation failure, one of the failure modes linked to RIA, are defined by peak fuel enthalpy values that range from 205 to 270 cal/g. This wide variability presents challenges when attempting to generalize criteria for modified-Zr alloy claddings with superior oxidation resistance. Therefore, it may be more relevant to apply failure criteria based on embrittlement mechanisms, such as oxygen concentration in the β-Zr phase. This study aimed to assess the failure based on both peak fuel enthalpy and cladding embrittlement by analyzing previous NSRR experiments conducted with conventional materials using the RANNS fuel performance code. The findings suggest that the failure criteria associated with cladding embrittlement can provide a rational evaluation of failure behavior compared to the existing criterion based on peak fuel enthalpy. The local failure criterion leading to the formation of through-wall cracks during quenching is consistent with Chung’s proposal (NUREG/CR-1344): β-Zr thickness of ≤ 0.9 wt% oxygen is less than 0.1 mm, and this corresponds to approximately 35 % BJ-ECR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114222"},"PeriodicalIF":1.9,"publicationDate":"2025-06-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144297611","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}

引用次数: 0

Position control of water hydraulic high-speed on-off valves-controlled cylinder of water hydraulic manipulator 液压高速开关阀控液压机械手缸的位置控制

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-06-14 DOI: 10.1016/j.nucengdes.2025.114235

Xing Yang, Defa Wu, Heng Gao, Yinshui Liu

{"title":"Position control of water hydraulic high-speed on-off valves-controlled cylinder of water hydraulic manipulator","authors":"Xing Yang, Defa Wu, Heng Gao, Yinshui Liu","doi":"10.1016/j.nucengdes.2025.114235","DOIUrl":"10.1016/j.nucengdes.2025.114235","url":null,"abstract":"<div><div>Water hydraulic systems offer the advantage of minimizing pollution, rendering them highly compatible with nuclear radiation environments. In this study, a 7 degrees of freedom (DOF) water hydraulic manipulator, based on seven water hydraulic high-speed on–off valves (HSVs)-controlled cylinder systems, is developed for operations in pollution-free environments. A co-simulation model for HSVs-controlled cylinder is established by AMESim, Maxwell, and Simulink. To improve the precision of manipulator joint control, a control strategy involving double voltage and variable frequency (DV+VF) for HSVs is introduced to ensure optimized flow output at any duty cycles. This approach effectively mitigates the decrease in position control accuracy caused by dead or saturated zones. Addressing the asymmetry within water cylinders, the DV+VF and double sliding mode control (DV+VF+DSMC) is proposed and implemented to achieve precise position tracking of the joint. Experimental results showcase that the displacement error of DV+VF+DSMC is within 0.5 mm while exhibiting stronger robustness. Furthermore, to simulate the working performance in high-pressure and pollution-free environments, the water hydraulic manipulator is tested within a high-pressure simulation device. The results indicate smooth and flexible movement for each manipulator joint, affirming its effectiveness in high-pressure and pollution-free environments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114235"},"PeriodicalIF":1.9,"publicationDate":"2025-06-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144279686","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}

引用次数: 0