Journal of Nuclear Materials最新文献

{"title":"The failure mechanism of nuclear-grade 316H protective oxide layer in long-term supercritical CO2 Corrosion","authors":"Zhihao Wang ,&nbsp;Jichun Zou ,&nbsp;Minghao Wang ,&nbsp;Shen Li ,&nbsp;Dequan Peng ,&nbsp;Shuai Chen ,&nbsp;Wanhuan Yang ,&nbsp;Weihua Zhong ,&nbsp;Wen Yang","doi":"10.1016/j.jnucmat.2025.155803","DOIUrl":"10.1016/j.jnucmat.2025.155803","url":null,"abstract":"<div><div>An experimental investigation was conducted to evaluate the corrosion behavior of nuclear-grade 316H austenitic stainless steel in supercritical CO₂ (S-CO₂) at 500 °C and 25 MPa over 6400 h. The analysis of the results indicated that, during the initial stage of corrosion, a protective Cr₂O₃ layer formed on the material's surface, exhibiting significant corrosion resistance. However, a sharp escalation in weight gain (13.4 × increase) was observed after 3200 h, accompanied by the formation of a dual-layered oxide structure (Fe₃O₄/FeCr₂O₄), indicating the failure of the Cr₂O₃ layer. This study elucidates the intrinsic chemical failure (InCF) mechanism of Cr₂O₃ under prolonged S-CO₂ exposure, providing critical insights for material selection in advanced nuclear systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"611 ","pages":"Article 155803"},"PeriodicalIF":2.8,"publicationDate":"2025-04-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143821048","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"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 critical analysis of U-Pu-Zr phase transitions using calorimetric, microstructural, and phase equilibria data","authors":"Scott Middlemas,&nbsp;Cynthia Adkins","doi":"10.1016/j.jnucmat.2025.155778","DOIUrl":"10.1016/j.jnucmat.2025.155778","url":null,"abstract":"<div><div>Metallic fuels consisting primarily of uranium, plutonium, and zirconium (U-Pu-Zr) are a leading material candidate for fast-spectrum nuclear reactors. Early demonstration programs proved the principle of safe and efficient fast reactor operation, however there is still considerable uncertainty regarding the phase equilibria and microstructural evolution across the ternary composition space. Quantitative phase formation and identification measurements are scarce and often incomplete, with studies reporting either phase transition temperatures or phase identification data, but not both from the same specimens. In this study, we critically compared experimental and calculated phase transition data and correlated with the microstructure and phase characterization data of as-cast and annealed U-Pu-Zr alloys. Differential scanning calorimetry (DSC) was used to measure phase transitions in the subsolidus regions (723−948 K) of three ternary U-Pu-Zr alloys with similar plutonium concentrations but various U/Zr ratios. Due to sluggish kinetics and narrow ranges of phase stability, complex peaks required the use of a Frazier-Suzuki peak fitting algorithm to deconvolute and calculate transition peak temperatures and enthalpies. We also identified trends of phase transition behavior by critically comparing our DSC data with previous phase transition measurements as well as historical and calculated phase equilibrium diagrams. This provides a critical approach for benchmarking and assessing the quality of new U-Pu-Zr phase equilibria data prior to its incorporation into nuclear material databases.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"612 ","pages":"Article 155778"},"PeriodicalIF":2.8,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143837761","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Low-temperature sintering of (U,Pu)O2 MOX in mild oxidative conditions","authors":"Jacobus Boshoven ,&nbsp;Jean-François Vigier ,&nbsp;Philipp Pöml ,&nbsp;Abibatou Ndiaye ,&nbsp;Bertrand Morel ,&nbsp;Rudy J.M. Konings ,&nbsp;Karin Popa ,&nbsp;Marco Cologna","doi":"10.1016/j.jnucmat.2025.155800","DOIUrl":"10.1016/j.jnucmat.2025.155800","url":null,"abstract":"<div><div>We compare typical reductive sintering conditions for U, Pu mixed oxides (4 h at 1700°C in Ar/6% H₂ + 1200 ppm H₂O) with lower temperature and mildly oxidative conditions (2 h at 1200°C in CO/CO₂ = 1/9) and report on the resulting microstructures and homogeneity. We show that lower temperature and mildly oxidative conditions, without cover gas change, can give close-to stoichiometric, crack-free MOX pellets with a relative density of ∼ 95%, and we propose ways to improve the homogenisations of PuO₂ and UO₂ and increase the grain size.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155800"},"PeriodicalIF":2.8,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143783935","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"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":"Development of the Technological Process for the IGR Reactor's Highly-Enriched Irradiated Uranium-Graphite Fuel Immobilization","authors":"Kuanysh K. Samarkhanov ,&nbsp;Yuliya Yu. Baklanova ,&nbsp;Olga S. Bukina ,&nbsp;Viktor V. Baklanov ,&nbsp;Yerbolat T. Koyanbayev ,&nbsp;Ivan M. Kukushkin ,&nbsp;Igor M. Bolshinsky ,&nbsp;Kenneth J. Bateman","doi":"10.1016/j.jnucmat.2025.155801","DOIUrl":"10.1016/j.jnucmat.2025.155801","url":null,"abstract":"<div><div>The immobilization of irradiated highly enriched uranium (HEU) fuel is a critical component of nuclear waste management and non-proliferation efforts. In Kazakhstan, at National Nuclear Center of the Republic of Kazakhstan special attention is given to managing legacy HEU fuel from research reactors. One such case involves the IGR research reactor, whose first core containing irradiated HEU uranium-graphite fuel was operated from 1961 to 1966 and removed following reactor modernization. This fuel now requires a reliable and secure immobilization strategy.</div><div>This paper presents the development of a technological process for immobilizing this fuel to reduce its enrichment to below 20% in terms of ²³⁵U content. The proposed method involves down-blending irradiated HEU fuel with depleted uranium, followed by encapsulation in a Portland cement matrix. Full-scale experiments were conducted to assess the uniformity of uranium distribution within the matrix.</div><div>The results confirm the effectiveness of this approach, ensuring reliable immobilization of fuel in accordance with international requirements, including IAEA standards and Kazakhstan's regulatory framework. These findings contribute to the broader effort of adapting immobilization strategies for the safe management of spent fuel from research reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155801"},"PeriodicalIF":2.8,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143783936","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"On the degradation of Young's modulus of irradiated U-10Mo","authors":"Chaoyue Jin ,&nbsp;Zhexiao Xie ,&nbsp;Luning Chen ,&nbsp;Xingdi Chen ,&nbsp;Jing Zhang ,&nbsp;Shurong Ding ,&nbsp;Xiaobin Jian","doi":"10.1016/j.jnucmat.2025.155799","DOIUrl":"10.1016/j.jnucmat.2025.155799","url":null,"abstract":"<div><div>Theoretical analysis of the four-point bending experimental results from the reference has demonstrated that the effective Young's modulus of heavily-irradiated U-10Mo fuel undergoes a significant reduction. However, the underlying mechanisms need to be fully elucidated. In this study, the irradiation-induced thermo-mechanical coupling behaviors of monolithic fuel plates are first numerically investigated by employing the fuel skeleton creep-based volumetric growth strain model and the porosity-related macroscale creep rate model for the contained U-10Mo fuel foils. The predicted average thicknesses for the bending specimens from several fuel plates align well with the experimental measurements, validating the adopted models, algorithms and the obtained macroscale porosity values for irradiated U-10Mo fuel. The values of effective Young's modulus of U-10Mo fuel after different levels of irradiation are identified through the subsequent direct simulations of the four-point bending tests, with the numerically acquired macroscale mechanical responses of irradiated U-10Mo specimens matching the experimental data. After eliminating the effects of fuel porosity, it is found that the values of Young's modulus of dense U-10Mo fuel skeleton decrease with increasing fission density or macroscale porosity, thereby becoming a primary contributor to the degradation of the effective Young's modulus of irradiated U-10Mo fuel. Furthermore, mathematical models for the Young's modulus of irradiated U-10Mo fuel skeleton at room temperature are developed as functions of fission density and macroscale porosity, respectively. The predicted results indicate that the von Mises stress will significantly decrease and the equivalent creep strains might have a distinct increase, when the degradation of Young's modulus of fuel skeleton is incorporated. This work provides a foundation for the high-precise modeling of the irradiation-induced thermo-mechanical behaviors of the U-10Mo-based fuel elements or assemblies.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155799"},"PeriodicalIF":2.8,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143814979","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Effects of pre-existing vacancy-type dislocation loop on the irradiation resistance in FeNiCoCrCu high-entropy alloy","authors":"Weidong Song ,&nbsp;Zhonghao Huo ,&nbsp;Lijun Xiao ,&nbsp;Lifang Wang ,&nbsp;Jun Chen ,&nbsp;Meizhen Xiang","doi":"10.1016/j.jnucmat.2025.155797","DOIUrl":"10.1016/j.jnucmat.2025.155797","url":null,"abstract":"<div><div>Several high-entropy alloys (HEAs) show considerable promise as structural materials for nuclear energy applications, owing to their exceptional mechanical properties and radiation resistance. However, there is limited understanding of how pre-existing dislocation loops in these HEAs influence their radiation resistance. This study employs molecular dynamics (MD) simulations to investigate the influence of pre-existing dislocation loops on the irradiation resistance of FeNiCoCrCu HEA, focusing on the evolution of point defects, the formation of defect clusters, and the interactions between dislocation loops and point defects during irradiation process. Consequently, the interaction between irradiation-induced point defects and pre-existing dislocation loops leads to an increase in the number of point defects and defect clusters. This is attributed to the reduction in formation energy of point defects by the dislocation loop, which promotes their generation and alters their distribution, thereby inhibiting recombination between them. Point defects migrate toward the dislocation loop under stress field interactions, with the loop exhibiting preferential absorption of interstitial atoms over vacancies, serving as predominant sinks for defect accumulation. Furthermore, the presence of dislocation loops mitigates elemental segregation. During irradiation, dislocation loops absorb vacancies via positive climb and interstitials via negative climb. Meanwhile, the position and shape of the dislocation loop undergo changes, and its length increases. Notably, the FeNiCoCrCu HEA demonstrates enhanced resistance to pre-existing vacancy loop interactions compared to pure Ni, as evidenced by fewer irradiation-induced defects and reduced dislocation loop evolution post-irradiation. These findings elucidate the intricate interplay of defect dynamics in irradiated HEAs and provide critical insights for designing radiation-tolerant HEA systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155797"},"PeriodicalIF":2.8,"publicationDate":"2025-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143783938","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Corrosion behavior and mechanical properties of SiC/SiC composite joints with Y2O3-Al2O3-SiO2 interlayer under high-temperature steam environments at 1200 °C","authors":"Shaobo Yang ,&nbsp;Chenxi Liang ,&nbsp;Jiali Li ,&nbsp;Yujie Ma ,&nbsp;Sijie Kou ,&nbsp;Juanli Deng ,&nbsp;Bo Chen ,&nbsp;Shangwu Fan","doi":"10.1016/j.jnucmat.2025.155798","DOIUrl":"10.1016/j.jnucmat.2025.155798","url":null,"abstract":"<div><div>The study investigated the corrosion behavior and mechanical performance of SiC/SiC composite joints with Y₂O₃-Al₂O₃-SiO₂ (YAS) interlayers under high-temperature steam environments at 1200 °C. Under low-flow conditions, partial disruption of Si-O and Al-O bonds in the YAS glass network reduced crosslinking, forming an aluminosilicate protective layer that inhibited further corrosion. Prolonged exposure led to Y³⁺ migration and accumulation, resulting in Y₂Si₂O₇ precipitation and growth. High-flow conditions caused a thinner glass layer, continuous longitudinal cracks, and more severe erosion and dissolution of the YAS glass due to higher steam velocity. Despite these degradations, the joints exhibited satisfactory performance, maintaining shear strengths of about 40 ± 2 MPa after 15 h of low-flow exposure and about 36 ± 5 MPa after 5 h of high-flow exposure. These findings demonstrate that YAS interlayers provide excellent corrosion resistance and mechanical stability as a sealant for nuclear-grade SiC/SiC.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155798"},"PeriodicalIF":2.8,"publicationDate":"2025-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768652","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Development of novel multi-element low-activation Fe-based alloys for nuclear and fusion reactor applications","authors":"Kazuyuki Furuya ,&nbsp;Koichi Tsuchiya ,&nbsp;Eiichi Wakai ,&nbsp;Elango Chandiran ,&nbsp;Bikash Tripathy ,&nbsp;Masami Ando ,&nbsp;Takaharu Kamada ,&nbsp;Hiroyuki Noto","doi":"10.1016/j.jnucmat.2025.155772","DOIUrl":"10.1016/j.jnucmat.2025.155772","url":null,"abstract":"<div><div>Microstructures, mechanical properties and irradiation hardening of novel multi-element iron-based alloys (Fe-(10 and 20)Mn-15Cr-2.0Al-0.7V-0.5C (at %)) were investigated. The alloys do not contain high activation elements, such as, Co, Ni and Mo. The alloy samples were hot-rolled at 1323 K and air-cooled, followed by heat treatment at 1073 K for 0.5 h and quenching in to water. After the heat-treatment, the Fe-10Mn-15Cr-2.0Al-0.7V-0.5C (10Mn) sample consisted mainly of body-centered cubic (BCC) structure with two distinct microstructures, i.e., fine lath-martensite-like structures and recrystallized grains. Meanwhile, the Fe-20Mn-15Cr-2.0Al-0.7V-0.5C (20Mn) sample were a mixture of fine lath-martensite like BCC phase and face-centered cubic (FCC) phases. The 10Mn sample exhibits very high tensile strength of 960 MPa but low elongation, while the 20Mn sample exhibits lower tensile strength of 620 MPa but much improved elongation over 60 %. The samples were simultaneously triple-irradiated with 10.5 MeV Fe³⁺ ions, 1.05 MeV He⁺ ions and 0.38 MeV H⁺ ions to a depth of 1 μm from the sample surface. The irradiation hardening in average was only about 1.5 GPa in the alloys irradiated with 10.5 MeV Fe³⁺ ions up to 30 dpa at 573 K at the damage peak, measured by nano-indentation. The irradiation hardening resistance of the alloys was better than that of other fusion structural materials and fission reactor pressure vessel steels. Combined analysis with electron-backscattered diffraction and nanoindentation revealed that the irradiation hardening is less significant in lath BCC phase than in recrystallized BCC (10Mn) and in FCC (20Mn). These results suggest that the alloys with good combination of irradiation resistance and mechanical properties can be developed by further tailoring the phase stability of the alloys and combining the high-entropy effects, aiming for the application for components in nuclear reactors, fusion reactors and high-power large accelerator facilities.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155772"},"PeriodicalIF":2.8,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143785554","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Impact of helium ion implantation on deuterium plasma induced microstructure evolution and deuterium retention in tungsten","authors":"Honghui Zhang ,&nbsp;Tongjun Xia ,&nbsp;Yongzhi Shi ,&nbsp;Zhengyu Jiang ,&nbsp;Xingyu Ren ,&nbsp;Lisha Liang ,&nbsp;Kaigui Zhu","doi":"10.1016/j.jnucmat.2025.155794","DOIUrl":"10.1016/j.jnucmat.2025.155794","url":null,"abstract":"<div><div>Surface blistering and internal microstructure evolutions as well as deuterium retention in tungsten with helium ion implanted followed by deuterium plasma exposure were investigated. The helium ion implantation was taken with 40 keV with a flux of 1.6 × 10¹⁷ He⁺/(m²s) to a fluence of 6.0 × 10²⁰ He⁺/m² at room temperature. The following deuterium plasma exposure was taken with a flux of 5.96 × 10¹⁹ D/(m²s) at a bias of 100 eV at 340 K. The deuterium plasma exposure was designed with two different durations. One is about 19 h (h) which corresponds a fluence of 4.07 × 10²⁴ D/m², while another is nearly 96 h corresponds a fluence of 2.06 × 10²⁵ D/m². The helium ion implantation itself did not induce surface blister nor detectable internal helium bubble. After subsequent deuterium exposure of 19 h, dense surface blisters appeared on the reference tungsten, while no blister was formed on the helium implanted tungsten, indicating the helium ion implantation can efficiently suppress the surface blistering. However, when the deuterium irradiation time was increased up to 96 h, sparse deuterium blisters appeared on the surface of the helium ion pre-implanted W, indicating D could pass through the helium implantation layer as the exposure time was long enough. TEM results revealed that no bubble can be observed in the reference tungsten only exposed to deuterium plasma, while bubbles can be confirmed in the helium ion pre-implanted tungsten after deuterium irradiation, suggesting that the growth of helium bubbles can be enhanced by the subsequent deuterium plasma exposure. For the deuterium plasma exposure with 19 h, the total deuterium retention in the helium ion pre-implanted tungsten was three times that of the reference tungsten, indicating the helium ion implantation could increase the deuterium retention in tungsten.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155794"},"PeriodicalIF":2.8,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143777488","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Impact of calcium and pH on ISG alteration at basic pH: Mechanism of formation and transport properties of the gel layer","authors":"Benjamin Cagnon ,&nbsp;Stéphane Gin ,&nbsp;Martiane Cabié ,&nbsp;Damien Daval","doi":"10.1016/j.jnucmat.2025.155796","DOIUrl":"10.1016/j.jnucmat.2025.155796","url":null,"abstract":"<div><div>The dissolution of International Simple Glass (ISG) was investigated at 90 °C, elevated concentration of dissolved silica and in the presence of calcium, with a specific emphasis on basic pH conditions. The leaching solution was labelled with ²⁹Si, ¹⁸O and ⁴⁴Ca in part of the experiments to elucidate the dissolution mechanisms. Based on the isotopic signatures of the gel layer analyzed using Time-of-Flight Secondary Ion Mass Spectrometry (ToF-SIMS), it was concluded that oxygen atoms mostly originate from the solution for all investigated conditions, while silicon atoms almost exclusively originate from the glass. A negative correlation was found between the initial concentration of calcium in solution and the gel layer thickness, suggesting either the formation of a passivating (Si, Ca)-rich layer, a catalytic effect of Ca on the gel densification or a combination of both. In addition, the pH-dependence of the diffusion coefficient of B within the gel was found to be stronger in the basic pH range than in the acidic pH range, which was suggested to originate from the change in coordination of B species at pH₉₀ °_C ∼ 8.5. Overall, these results suggest that in a (Ca, Si)-rich solution at basic pH, the durability of ISG is stronger than previously thought, as the diffusion coefficient of B under such conditions are lower than expected based on literature.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155796"},"PeriodicalIF":2.8,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768659","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}