Annals of Nuclear Energy最新文献

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Numerical modeling and validation for transport behaviors of tracer gas in the European tracer experiment (ETEX-1) 欧洲示踪实验(ETEX-1)中示踪气体输运行为的数值模拟与验证
IF 2.3 3区 工程技术
Annals of Nuclear Energy Pub Date : 2025-10-18 DOI: 10.1016/j.anucene.2025.111953
Zhenhui Ma, Xiuhuan Tang, Longbo Liu, Lihong Bao, Baosheng Wang, Tengyue Ma, Pan Hu, Yonggang Zhangsun, Chunlei Su
{"title":"Numerical modeling and validation for transport behaviors of tracer gas in the European tracer experiment (ETEX-1)","authors":"Zhenhui Ma,&nbsp;Xiuhuan Tang,&nbsp;Longbo Liu,&nbsp;Lihong Bao,&nbsp;Baosheng Wang,&nbsp;Tengyue Ma,&nbsp;Pan Hu,&nbsp;Yonggang Zhangsun,&nbsp;Chunlei Su","doi":"10.1016/j.anucene.2025.111953","DOIUrl":"10.1016/j.anucene.2025.111953","url":null,"abstract":"<div><div>The European Tracer Experiment (ETEX-1) was conducted across Europe by European Commission, World Meteorological Organization and International Atomic Energy Agency jointly on October 26th, 1994. In the experiment, PMCH (Perflouro-Methyl-Cyclo-Hexane) was adopted as tracer gas, and it was released from Monterfil, France for nearly 24 hours. Totally 168 observation stations for PMCH were settled in different countries in Europe to observe time-varying concentration for arriving tracer gas. This study established a numerical method for simulating tracer gas transport using the meteorology-air quality coupling model WRF-CMAQ and applied this method to simulate tracer gas transport during ETEX-1. On the basis of simulation results for PMCH concentration in atmosphere, transport characteristics can be analyzed in detail. According to simulation results, 72 hours after release, tracer gas could influence most part of central and southeast Europe. Furthermore, gas concentrations at different times from simulation results were compared with data observed from representative observation stations. Finally, a comprehensive statistical analysis was performed to evaluate the model’s performance in meso-scale atmospheric transport simulation. According to analysis results for statistical indicators, although the value of FB was slightly lower than the recommended distribution range (FB = −0.308, recommended distribution range: −0.3 ∼ 0.3), Correlational Coefficient (r = 0.275, recommended distribution range: −1.0 ∼ 1.0) and NMSE (NMSE = 1.26, recommended distribution range: &lt;4.0) were both in the acceptable distribution range, which means the present model is comparable with other models, even show better performance in analysis for some statistical indicators.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111953"},"PeriodicalIF":2.3,"publicationDate":"2025-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322536","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
Fuel performance modeling of used nuclear fuel transfer between storage pools and its impact on dry storage 乏燃料在贮存池间转移的燃料性能建模及其对干贮存的影响
IF 2.3 3区 工程技术
Annals of Nuclear Energy Pub Date : 2025-10-18 DOI: 10.1016/j.anucene.2025.111934
Piotr Konarski, Cedric Cozzo, Grigori Khvostov, Hakim Ferroukhi
{"title":"Fuel performance modeling of used nuclear fuel transfer between storage pools and its impact on dry storage","authors":"Piotr Konarski,&nbsp;Cedric Cozzo,&nbsp;Grigori Khvostov,&nbsp;Hakim Ferroukhi","doi":"10.1016/j.anucene.2025.111934","DOIUrl":"10.1016/j.anucene.2025.111934","url":null,"abstract":"<div><div>The transfer of fuel in dry casks between the storage pools at the Swiss nuclear power plant Gösgen is known as the Wet-To-Dry-To-Wet (W2D2W) transfer procedure. The process is associated with vacuum drying, transfer to an external storage pool and reflooding. The high cladding temperature and stress during the drying and transport in combination with quick reflooding can lead to precipitation of radial hydrides and deterioration of the cladding mechanical properties. In this report, the impact of the drying, transfer and reflooding periods on the concentration and distribution of radial hydrides is studied. The most limiting rods are identified based on the criteria proposed in this work and submitted to multiple scenarios considering various durations of the W2D2W stages. The results show the impact of different phases and reveal that the reflooding speed/rate has the biggest influence on the hydrogen behavior. The analysis of the simulated dry storage scenarios shows that the maximum temperature obtained at the end of drying governs the hydrogen migration into the liner.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111934"},"PeriodicalIF":2.3,"publicationDate":"2025-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322537","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
Optimisation of waste management from the remediation of uranium contaminated soils 铀污染土壤修复中废物管理的优化
IF 2.3 3区 工程技术
Annals of Nuclear Energy Pub Date : 2025-10-18 DOI: 10.1016/j.anucene.2025.111955
A. Álvarez, J.A. Suárez-Navarro, M.A.Lopez Ponte, V.M. Expósito-Suárez
{"title":"Optimisation of waste management from the remediation of uranium contaminated soils","authors":"A. Álvarez,&nbsp;J.A. Suárez-Navarro,&nbsp;M.A.Lopez Ponte,&nbsp;V.M. Expósito-Suárez","doi":"10.1016/j.anucene.2025.111955","DOIUrl":"10.1016/j.anucene.2025.111955","url":null,"abstract":"<div><div>Soil remediation at nuclear sites is necessary to protect ecosystems and public health. The design of remediation includes an appropriate measurement strategy to reduce economic and environmental costs. Field measurements of soils containing natural U are challenging due to the low energy and intensity of <sup>238</sup>U daughter gamma emissions. Due to self-absorption at the source, the detection of the <sup>234m</sup>Pa beta emission requires the establishment of an adequate thickness, making it difficult to measure in the field.</div><div>This paper describes the design of an automated measurement system for the in situ measurement soils contaminated with uranium. The system allows the radiological classification of soils according to their level of compliance with the criterion of unrestricted release of the site. Its use has the advantage of providing results faster and at a lower cost than laboratory sampling. The result is a real optimisation of radioactive waste management.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111955"},"PeriodicalIF":2.3,"publicationDate":"2025-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322484","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
Secondary side system design: A new calculation method for passive residual heat removal of liquid fuel molten salt reactors 二次侧系统设计:液体燃料熔盐堆被动余热排出的一种新的计算方法
IF 2.3 3区 工程技术
Annals of Nuclear Energy Pub Date : 2025-10-18 DOI: 10.1016/j.anucene.2025.111946
Shuaiyu Xue , Chong Zhou , Pinyan Huang , Yang Zou
{"title":"Secondary side system design: A new calculation method for passive residual heat removal of liquid fuel molten salt reactors","authors":"Shuaiyu Xue ,&nbsp;Chong Zhou ,&nbsp;Pinyan Huang ,&nbsp;Yang Zou","doi":"10.1016/j.anucene.2025.111946","DOIUrl":"10.1016/j.anucene.2025.111946","url":null,"abstract":"<div><div>The liquid molten salt reactor is distinguished by its liquid fuel, which results in the residual heat being dispersed throughout the circuit by flow. Consequently, the design calculation methods for natural circulation of the molten salt reactor diverge from traditional methods. To address the structural design calculations for this type of reactor, we propose a technique that balances the natural circulation pressure drop within the circuit with the heat transfer equilibrium of the heat exchangers between circuits. We develop the Molten-salt-reactor Passive-residual-heat-removal-system Design and Calculation Program (MPDCP) based on this approach. The results prove that the distributed residual heat within the primary circuit will reduce the heat dissipation efficiency. By using the design program, a set of passive residual heat removal systems that meet the design objectives can be obtained, effectively improving the heat dissipation efficiency by 19.1%, meeting the heat dissipation needs of the liquid fuel molten salt reactor.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111946"},"PeriodicalIF":2.3,"publicationDate":"2025-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322483","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
Feasibility Assessment of Thorium/Plutonium-Based Accident-Tolerant fuel in an Innovative fuel assembly design for the VVER-1200 reactor 基于钍/钚的耐事故燃料在VVER-1200反应堆创新燃料组件设计中的可行性评估
IF 2.3 3区 工程技术
Annals of Nuclear Energy Pub Date : 2025-10-18 DOI: 10.1016/j.anucene.2025.111954
Nassar Alnassar , Shlash A. Luhaib , Mohamed Y.M. Mohsen , Mostafa M.A. Khater , A. Abdelghafar Galahom
{"title":"Feasibility Assessment of Thorium/Plutonium-Based Accident-Tolerant fuel in an Innovative fuel assembly design for the VVER-1200 reactor","authors":"Nassar Alnassar ,&nbsp;Shlash A. Luhaib ,&nbsp;Mohamed Y.M. Mohsen ,&nbsp;Mostafa M.A. Khater ,&nbsp;A. Abdelghafar Galahom","doi":"10.1016/j.anucene.2025.111954","DOIUrl":"10.1016/j.anucene.2025.111954","url":null,"abstract":"<div><div>Radioactivity reduction and effective waste management remain critical challenges for the nuclear industry, affecting long-term safety and public acceptance. This study explores the feasibility of employing thorium–plutonium accident-tolerant fuels (ATFs) in a novel blanket–seed (BS) assembly for the VVER-1200 reactor. Comparative evaluations were conducted based on fuel burnup, thermal power distribution, key safety-related parameters, and radioactivity analysis. Also, the decay of spent fuel radioactivity during the cooling process was assessed for fuel stored in a pool containing borated water with a boron concentration of 600 pcm. Results indicate that all BS configurations achieve extended reactivity lifetimes (up to ∼ 1750 EFPDs), and enhanced neutron economy due to higher atomic densities in denser fuels. The BS assembly consumed a significant amount of the plutonium used in the seed rods. Compared to UO<sub>2</sub>, thorium-based fuels exhibit significantly reduced total actinide and non-actinide radioactivity, demonstrating a favorable trade-off for long-term waste management.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111954"},"PeriodicalIF":2.3,"publicationDate":"2025-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322535","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
Multi-attribute evaluation method of nuclear island plant overall layout schemes based on linguistic intuitionistic fuzzy VIKOR 基于语言直觉模糊VIKOR的核岛电厂总体布局方案多属性评价方法
IF 2.3 3区 工程技术
Annals of Nuclear Energy Pub Date : 2025-10-18 DOI: 10.1016/j.anucene.2025.111950
Dong Hao , JinCheng Su, YanFang Fan
{"title":"Multi-attribute evaluation method of nuclear island plant overall layout schemes based on linguistic intuitionistic fuzzy VIKOR","authors":"Dong Hao ,&nbsp;JinCheng Su,&nbsp;YanFang Fan","doi":"10.1016/j.anucene.2025.111950","DOIUrl":"10.1016/j.anucene.2025.111950","url":null,"abstract":"<div><div>This paper presents a novel method for evaluating nuclear island layout schemes using linguistic intuitionistic fuzzy sets and the VIKOR approach. Nuclear island layout involves multiple, often conflicting criteria such as safety, functionality, human factors, radiation protection, and cost. Traditional methods struggle to address uncertainty and complexity in such decisions. The proposed approach builds a structured evaluation index system, uses DEMATEL-ANP to determine indicator weights, and applies linguistic intuitionistic fuzzy numbers with VIKOR for multi-criteria decision-making. The method has been validated through the evaluation and decision-making of layout schemes for the ACP100 small modular reactor, demonstrating the method’s reliability and scientific basis for evaluating nuclear island layouts. The results highlight its capability to systematically manage complexity and uncertainty while balancing diverse design priorities.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111950"},"PeriodicalIF":2.3,"publicationDate":"2025-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322485","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
Comparative analysis of CANDU severe accident using CAISER 应用CAISER对CANDU严重事故进行对比分析
IF 2.3 3区 工程技术
Annals of Nuclear Energy Pub Date : 2025-10-17 DOI: 10.1016/j.anucene.2025.111931
Jun-young Kang Ph.D., Yong Mann Song, Dong Gun Son, Keun Sang Choi, Jong Yeob Jung, Jun Ho Bae, Byeonghee Lee
{"title":"Comparative analysis of CANDU severe accident using CAISER","authors":"Jun-young Kang Ph.D.,&nbsp;Yong Mann Song,&nbsp;Dong Gun Son,&nbsp;Keun Sang Choi,&nbsp;Jong Yeob Jung,&nbsp;Jun Ho Bae,&nbsp;Byeonghee Lee","doi":"10.1016/j.anucene.2025.111931","DOIUrl":"10.1016/j.anucene.2025.111931","url":null,"abstract":"<div><div>CAISER (CANDU Advanced Integrated SEveRe accident analysis) is the recently developed the system code to evaluate the severe accident of CANDU plant and models the CANDU reactor core, incorporating 380 fuel channels and 37 fuel pins within a Cartesian-coordinate node system. It enables detailed assessments of mass and temperature distributions inside fuel channels including core degradation and relocation. CAISER code is coupled with the reactor thermal–hydraulic module (MARS-KS), the ex-vessel containment module (CONTAIN) and source term module (SIRIUS) using dynamic linked libraries. Present study evaluated the reference scenario of severe accident from IAEA-CRP (TECDOC-1727), benchmarking severe accident simulation tools for CANDU applications. Calculated results by CAISER are compared with that of IAEA-CRP and MAAP-ISAAC v4.03 and are discussed from the in-calandria to the ex-calandria phenomena.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111931"},"PeriodicalIF":2.3,"publicationDate":"2025-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322488","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
Revisiting mini-max polynomial approximation method for nuclear fuel depletion calculation 再论核燃料耗竭计算的最小-最大多项式近似法
IF 2.3 3区 工程技术
Annals of Nuclear Energy Pub Date : 2025-10-17 DOI: 10.1016/j.anucene.2025.111948
Go Chiba , Kento Yamamoto , Hiroaki Nagano
{"title":"Revisiting mini-max polynomial approximation method for nuclear fuel depletion calculation","authors":"Go Chiba ,&nbsp;Kento Yamamoto ,&nbsp;Hiroaki Nagano","doi":"10.1016/j.anucene.2025.111948","DOIUrl":"10.1016/j.anucene.2025.111948","url":null,"abstract":"<div><div>Nuclear fuel depletion calculations with detailed nuclide transmutation chains require specific numerical methods and the Chebyshev rational approximation method (CRAM) has been widely used. The mini-max polynomial approximation (MMPA) method is also for fuel depletion calculations and has several advantages over CRAM. The original MMPA coefficients are determined to minimize polynomial approximation errors over an entire range of a variable. In the present paper, relation between the polynomial approximation errors and reproduction errors of nuclide number densities (ND) is carefully investigated, and it is found that the MMPA coefficients which minimize approximation errors in a specific range of the variable can reduce the reproduction errors of ND. Reference NDs are reproduced within 1% differences with the new MMPA coefficients with the 8th order for fuel depletion calculations of PWR-simulated pincells.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111948"},"PeriodicalIF":2.3,"publicationDate":"2025-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322486","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
Analysis of the effect of LFR bending fuel assembly on thermal–hydraulic characteristics LFR弯曲燃料组件对热工特性的影响分析
IF 2.3 3区 工程技术
Annals of Nuclear Energy Pub Date : 2025-10-17 DOI: 10.1016/j.anucene.2025.111933
JiTong Sun , Mei Huang , YaoDi Li , AoNan Li , YanTing Cheng
{"title":"Analysis of the effect of LFR bending fuel assembly on thermal–hydraulic characteristics","authors":"JiTong Sun ,&nbsp;Mei Huang ,&nbsp;YaoDi Li ,&nbsp;AoNan Li ,&nbsp;YanTing Cheng","doi":"10.1016/j.anucene.2025.111933","DOIUrl":"10.1016/j.anucene.2025.111933","url":null,"abstract":"<div><div>To investigate the effect of fuel rod bending on coolant flow and heat transfer in lead-cooled fast reactors, the CFD models of C-shaped bending fuel assembly are established in this paper. Results show that fuel rod bending reduces heat transfer efficiency by 2.1 %, 5.4 %, and 7.4 % under Cases 1–3, respectively. Compared with the normal condition, the coolant velocity in corner and edge subchannels on the bending side decreases, with the maximum temperature rises reaching 11.12 K and 13.5 K under Case 3, respectively. On the bending dorsal side, the coolant velocity in both corner and edge subchannels exceeds the LFR design limit of 2 m/s under Cases 2 and 3. Bending also amplifies the deformation of fuel rods under fluid load, causing stress concentration on the 15th corner rod, the maximum deformation rises by 0.069 mm and the von Mises stress by 2.04 MPa under Case 3.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111933"},"PeriodicalIF":2.3,"publicationDate":"2025-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322490","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
An adaptive three-dimensional model for thermal hydraulics analysis of a PWR steam generator 一种用于压水堆蒸汽发生器热水力分析的自适应三维模型
IF 2.3 3区 工程技术
Annals of Nuclear Energy Pub Date : 2025-10-17 DOI: 10.1016/j.anucene.2025.111922
Yung Suk Nam , Han Young Yoon
{"title":"An adaptive three-dimensional model for thermal hydraulics analysis of a PWR steam generator","authors":"Yung Suk Nam ,&nbsp;Han Young Yoon","doi":"10.1016/j.anucene.2025.111922","DOIUrl":"10.1016/j.anucene.2025.111922","url":null,"abstract":"<div><div>Pressurized Water Reactor (PWR) steam generators (SGs) feature complex internal structures, including numerous heat exchanger tubes and support components. Accurate modeling of these internal structures is essential for reliable thermal–hydraulic analysis. However, conventional modeling approaches for SG thermal hydraulic analysis often require significant effort and time, limiting efficiency and flexibility. In this paper, we propose an adaptive three-dimensional (3D) modeling approach for the internal structures of PWR SGs, enabling comprehensive and efficient thermal–hydraulic analysis applicable to various designs. The 3D mesh encompasses all critical regions of the SG, including the U-tube heat exchanger, riser, downcomer, and steam dome. A dedicated mesh model accurately representing the shape of the U-tube bundle was developed, with the total tube length calculated from the model differing by less than 1% compared to design specifications. The CUPID-SG code (CUPID for SG application), based on the two-fluid model, is employed to perform the thermal–hydraulic analysis using the proposed 3D mesh. CUPID-SG incorporates primary coolant, U-tube heat conduction, and secondary coolant two-phase flow models, providing a comprehensive framework for simulation. The effectiveness of the proposed approach is demonstrated through the thermal–hydraulic analysis of the APR1400 SG, validating its capability for accurate and adaptable modeling of PWR SGs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111922"},"PeriodicalIF":2.3,"publicationDate":"2025-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322482","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
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