Journal of Nuclear Materials最新文献_第9页

Assessment of structural stability and leaching characteristics of phosphate-based geopolymer waste form containing radioactive spent ion exchange resins

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-02 DOI: 10.1016/j.jnucmat.2025.155671

Byoungkwan Kim , Younglim Shin , Solmoi Park , Brant Walkley , Seokyoung Oh , Jaehyuk Kang , Wooyong Um

{"title":"Assessment of structural stability and leaching characteristics of phosphate-based geopolymer waste form containing radioactive spent ion exchange resins","authors":"Byoungkwan Kim , Younglim Shin , Solmoi Park , Brant Walkley , Seokyoung Oh , Jaehyuk Kang , Wooyong Um","doi":"10.1016/j.jnucmat.2025.155671","DOIUrl":"10.1016/j.jnucmat.2025.155671","url":null,"abstract":"<div><div>The use of hydrated waste forms, such as cement, for immobilizing radioactive spent ion-exchange resins (IERs) is unsuitable due to low waste loading, high leaching of radionuclides, and poor durability. Here, simulant spent IERs were immobilized by a phosphate-based geopolymer (P-GP) for the first time. The 7-day compressive strength of the P-GP waste form was inversely proportional to waste loading because the pore size of the P-GP waste form increased with increasing waste loading. The P-GP waste forms with 40 wt% spent IERs satisfied all of South Korea's waste acceptance criteria for compressive strength, thermal cycling, water immersion, and gamma irradiation tests. The leaching behaviors of Co, Cs, and Sr differed from those of alkali-activated materials, but the leaching index exceeded the criterion value of 6.0. The leaching mechanism was governed by the combination of surface wash-off and diffusion or solely diffusion. The P-GP waste form could play as a primary physical barrier against releasing radionuclides. In addition, the geopolymer waste form did not undergo significant structural changes after waste acceptance criteria tests, indicating that it can efficiently immobilize spent IERs. Our findings can contribute new insights into efficient waste form materials for immobilizing radioactive spent IERs.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155671"},"PeriodicalIF":2.8,"publicationDate":"2025-02-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143237313","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}

引用次数: 0

Exploring the influence of alloying elements (Zr, Si, and V) on the corrosion behavior of FeCoNiCr-based high entropy alloys in supercritical water environment

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155605

Yuxuan Xia , Shanliang Zhou , Zihang Liang , Xi Huang , Qi Zhao , Huiwang Huang , Haidan Zhao , Yilei Zhan , Changfu Wang , Yujun Xie

引用次数: 0

Comparative analysis of the long-term strength of Russian ferritic-martensitic reactor steels

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155575

N.V. Kataeva , V.V. Sagaradze , V.A. Zavalishin , K.A. Kozlov , V.A. Sirosh , M.V. Leont'eva-Smirnova , A.A. Nikitina

{"title":"Comparative analysis of the long-term strength of Russian ferritic-martensitic reactor steels","authors":"N.V. Kataeva , V.V. Sagaradze , V.A. Zavalishin , K.A. Kozlov , V.A. Sirosh , M.V. Leont'eva-Smirnova , A.A. Nikitina","doi":"10.1016/j.jnucmat.2024.155575","DOIUrl":"10.1016/j.jnucmat.2024.155575","url":null,"abstract":"<div><div>The paper presents the results of long-term high-temperature creep tests of Russian reactor steels with ferritic-martensitic structure (the duration of some measurements exceeded 8 years). In the current study, the structural-phase transformations, characteristics of creep and long-term strength at 650 °C, 670 °C, and 700 °C under 60–140 MPa in oxide-free and oxide containing steels were determined. The creep tests were performed on specially designed transverse micro-specimens prepared from fuel elements cladding used in the fast-neutron reactor. The creep velocity of the ferritic-martensitic reactor steels was established to be specified by resistance of lath martensite and ferrite structures to diffusion processes of return and recrystallization. The most heat-resistant oxide-free steel contains the largest amount of refractory elements and carbides. The best heat resistance was observed for the steel hardened with thermal-resistant yttrium-titanium nanooxides. The samples made of this steel demonstrated one order less creep velocity at 700 °C under 100 MPa and 100-fold time to fracture in comparison with the oxide-free reactor steels.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155575"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170554","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}

引用次数: 0

Chemical characterization of uniform and localized oxides formed on Zr-2.5 (wt. %) Nb alloy

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155546

Akanksha Samanta , Suparna Banerjee , Rekha Rao , V. Srikanth , Y. Sunitha , Santu Kaity , J.V. Ramana , A.K. Tyagi

{"title":"Chemical characterization of uniform and localized oxides formed on Zr-2.5 (wt. %) Nb alloy","authors":"Akanksha Samanta , Suparna Banerjee , Rekha Rao , V. Srikanth , Y. Sunitha , Santu Kaity , J.V. Ramana , A.K. Tyagi","doi":"10.1016/j.jnucmat.2024.155546","DOIUrl":"10.1016/j.jnucmat.2024.155546","url":null,"abstract":"<div><div>Uniform and localized oxide morphologies formed on Zr-2.5(wt.%) Nb alloy in water and steam environments have been studied using electron microscopy techniques such as SEM and EPMA. The uniform oxidation was characterized by a compact and thin oxide layer. In contrast, the nodules showed localized thicker oxide growth with cracked and porous morphology. Significant segregation of Nb at the surface of the oxide nodule was observed, leading to destabilization and spallation of the oxide. The chemical states of Zr, Nb, and O in the two types of oxides were analyzed by XPS. The atomic composition of oxides in terms of O and M fractions was evaluated by p-EBS. The study revealed the oxide nodule to be richer in oxygen in comparison to the surrounding uniform oxide. The phase characteristics of these oxides were studied by Raman spectroscopy. The effect of initial surface in-homogeneities on nodule nucleation has been discussed. Polishing made the surface free of initial in-homogeneities and resulted in uniform oxidation on the surface.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155546"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170558","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}

引用次数: 0

Fracture behavior of irradiation induced nanocrystalline UO2 studied by in-situ mechanical testing in transmission electron microscopy

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155571

Laura Hawkins , Ana Garcia Caraveo , David Frazer , Fabiola Cappia , Tianyi Chen , Collin Knight , Jeffrey J. Giglio , Tiankai Yao

{"title":"Fracture behavior of irradiation induced nanocrystalline UO2 studied by in-situ mechanical testing in transmission electron microscopy","authors":"Laura Hawkins , Ana Garcia Caraveo , David Frazer , Fabiola Cappia , Tianyi Chen , Collin Knight , Jeffrey J. Giglio , Tiankai Yao","doi":"10.1016/j.jnucmat.2024.155571","DOIUrl":"10.1016/j.jnucmat.2024.155571","url":null,"abstract":"<div><div>Uranium Dioxide (UO<sub>2</sub>) is widely used as a fuel in current light water reactors (LWRs). Upon accumulation of radiation damage, LWR UO<sub>2</sub> fuel pellets start to develop a different microstructure at the pellet periphery when fuel burnup exceeds 45–50 GWd/tHM. The resulting porous, nanocrystalline microstructure is one of the most prominent microstructural changes occurring in such fuel. Its fracture mechanisms, which causes fuel fine fragmentation, could impact safety limits when the cladding breaches. Direct measurements of these properties are challenging, therefore a surrogate obtained via ion irradiation can be used. In this study, multiple microcantilevers were fabricated by focused ion beam from both fresh UO<sub>2</sub> and UO<sub>2</sub> irradiated with 84 MeV Xe<sup>26+</sup> ions to a peak dose of 1357 displacements per atom (dpa). The irradiation produced a pseudo high burnup structure approximately 2 µm below the surface. In-situ nano-mechanical bending tests were conducted to investigate the fracture behavior and the effect of the surrogate UO<sub>2</sub> high burnup structure on local fracture properties. Fresh UO<sub>2</sub> fuel was observed to fracture in transgranular mode without nucleation or movement of dislocations. However, the Xe-irradiated nanocrystalline microcantilevers fractured along the grain boundaries, with no influence from the pre-existing micro-cracks in the microcantilever. Fracture toughness for this type of surrogate high burnup UO<sub>2</sub> structure is reported for the first time in literature. Both the fracture stress and toughness show degradation for UO<sub>2</sub> as a result of Xe-irradiation.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155571"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170561","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}

引用次数: 0

Self-ion-irradiation-induced grain formation in nickel

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155588

S.E. Donnelly , G. Greaves , F. Granberg , J. Sharp , A.H. Milston , J.A. Hinks , K. Nordlund

{"title":"Self-ion-irradiation-induced grain formation in nickel","authors":"S.E. Donnelly , G. Greaves , F. Granberg , J. Sharp , A.H. Milston , J.A. Hinks , K. Nordlund","doi":"10.1016/j.jnucmat.2024.155588","DOIUrl":"10.1016/j.jnucmat.2024.155588","url":null,"abstract":"<div><div>A series of experiments has been conducted in which thin foils containing large polycrystals of Ni (single crystals from the perspective of transmission electron microscopy) have been irradiated with 300 keV Ni ions at temperatures from 25° to 475°C. The aim was to examine the fundamental aspects of the build-up of extended defects in a “simple” system with no implantation of foreign species and without the likelihood of segregation, precipitation or formation of new phases. Experiments were carried out using the MIAMI-2 facility in which the development of radiation damage is observed (and recorded) whilst ion-irradiating in-situ in a transmission electron microscope. Surprisingly, all irradiations of the electrochemically-thinned foils of Ni resulted in the accumulation of dislocations to form low-angle grain boundaries such that single crystal material was converted into a series of grains, each typically less than 200 nm in width but generally more than 1 <em>µ</em>m in length with the long axis approximately parallel to the edge of the foil. The early stages of this process have been modelled using Molecular Dynamics simulations and an interpretation of this process of radiation-induced grain-boundary formation is discussed in terms of the coupled effects of irradiation, temperature and stress induced by the radiation damage. The stress arises due to swelling in the thin irradiated region of the jet-polished specimens (with a wedge-shaped radial cross section) which is constrained by deeper-lying unirradiated material. The position in which a grain boundary forms is determined by the interaction of glisssile dislocations with the stress induced by the radiation-damaged layer and that from a neighbouring boundary.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155588"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143170986","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}

引用次数: 0

Dissolution behavior of U3O8, FeUO4, and UO2-Zr-stainless steel system samples generated in an oxidative atmosphere in the presence of malonic acid

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155561

Ryutaro Tonna , Takayuki Sasaki , Yoshihiro Okamoto , Taishi Kobayashi

{"title":"Dissolution behavior of U3O8, FeUO4, and UO2-Zr-stainless steel system samples generated in an oxidative atmosphere in the presence of malonic acid","authors":"Ryutaro Tonna , Takayuki Sasaki , Yoshihiro Okamoto , Taishi Kobayashi","doi":"10.1016/j.jnucmat.2024.155561","DOIUrl":"10.1016/j.jnucmat.2024.155561","url":null,"abstract":"<div><div>Understanding the dissolution behavior of the UO<sub>2</sub>-Zr-stainless steel (SS) system in the presence of organic acids is crucial for the safe decommissioning of the Fukushima Daiichi Nuclear Power Plants, as well as for the future treatment toward deep geological disposal of fuel debris. The dissolution behavior of UO<sub>2</sub>-Zr-SS samples heated under oxidative conditions, along with the single uranium phases formed in the samples (U<sub>3</sub>O<sub>8</sub> and FeUO<sub>4</sub>)<sub>,</sub> was investigated through static leaching tests using malonic acid. Malonic acid promoted the dissolution of both U<sub>3</sub>O<sub>8</sub> and FeUO<sub>4</sub> solid phases owing to complex formation. The uranium concentrations of U<sub>3</sub>O<sub>8</sub> and FeUO<sub>4</sub> increased with malonic acid concentration and matched at steady state. In the absence of malonic acid, the U concentration of FeUO<sub>4</sub> was less than its solubility because of uranium adsorption on an iron hydrolysis solid phase. The uranium concentration after long immersion of the UO<sub>2</sub>-Zr-SS system samples could be explained by the behavior of U<sub>3</sub>O<sub>8</sub> and FeUO<sub>4</sub>. In the absence of malonic acid, U concentration decreased owing to adsorption onto the iron solid phase, similar to the behavior of FeUO<sub>4</sub>. In contrast, in the presence of malonic acid, U concentration was consistent with that observed for U<sub>3</sub>O<sub>8</sub> and FeUO<sub>4</sub>.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155561"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171602","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}

引用次数: 0

A model for trapping and re-solution regarding intra-granular bubbles in UO2, linked to atomic-scale simulations

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2024.155562

M. Vergani , M.W.D. Cooper , L. Noirot

{"title":"A model for trapping and re-solution regarding intra-granular bubbles in UO2, linked to atomic-scale simulations","authors":"M. Vergani , M.W.D. Cooper , L. Noirot","doi":"10.1016/j.jnucmat.2024.155562","DOIUrl":"10.1016/j.jnucmat.2024.155562","url":null,"abstract":"<div><div>In the literature, a clear definition of the irradiation re-solution frequency of gas from bubbles in the UO<sub>2</sub> fuel is absent. Moreover, for intra-granular bubbles, a detailed calculation of the cumulated displaced gas quantities in function of the distance from the radius of the bubble after a re-solution event has never been published. The assessment of these two elements is very useful if we want to increase the adherence of fission gas release codes to our present knowledge of the behavior of fission gases. Hence, we suggest to link the definition of the re-solution frequency to atomic-scale simulations. Furthermore, we present the cumulated displaced gas quantities obtained from Molecular Dynamics calculations, from which we have derived a re-solution profile that can be exploited to better consider the irradiation re-solution phenomenon inside Fission Gas Release codes. On top of that, we have built a new trapping/re-solution model for intra-granular bubbles linked to Molecular Dynamics simulations that can be easily incorporated into Fission Gas Release codes. We also check that the model is properly built through the comparison of the new model against a reference.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"605 ","pages":"Article 155562"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143171603","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}

引用次数: 0

Development of irradiation- and high-temperature resistant steels for fusion applications: Belgian contribution

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155611

D. Terentyev , O. Kachko , A. Puype , S. Valiyev , K. Iroc , A. Zinovev

{"title":"Development of irradiation- and high-temperature resistant steels for fusion applications: Belgian contribution","authors":"D. Terentyev , O. Kachko , A. Puype , S. Valiyev , K. Iroc , A. Zinovev","doi":"10.1016/j.jnucmat.2025.155611","DOIUrl":"10.1016/j.jnucmat.2025.155611","url":null,"abstract":"<div><div>In this work, we investigate alternative routes for the production of reduced activation ferritic-martensitic (RAFM) steels aiming to achieve specific improvements of their performance under fusion operational conditions. The latter impose at least two specific challenges: (i) low-temperature embrittlement (LTE) and (ii) high-temperature creep (HTC) deformation. In this work, we review the optimization routes attempted to alleviate the above noted challenges which are otherwise met in EUROFER97 steel. The development routes include: (i) reduction of manganese and carbon content coupled with alternation of other chemical elements and followed by quench & rolling procedures; (ii) alternation of spatial distribution and structural morphology of carbonitrides by varying carbon, vanadium and tantalum content based on thermodynamic computations and followed by thermo-mechanical treatment optimization; (iii) doping with zirconium/titanium and increase of tantalum content to improve ductility and toughness. The targeted enhanced performance is achieved without compromising strength and DBTT. The results of the baseline characterization including mechanical tests and microstructural characterization are presented. The contribution of the microstructural features constituting the ferritic martensitic steels into the tensile strength is analyzed based on existing mechanistic models and discussed to rationalize the improvements achieved.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155611"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154915","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}

引用次数: 0

Microstructure evolution in titanium carbide with different stoichiometry under 3 MeV Au2+ ion irradiation

IF 2.8 2区工程技术

Journal of Nuclear Materials Pub Date : 2025-02-01 DOI: 10.1016/j.jnucmat.2025.155609

Jinyu Shi , Yiming Lei , Chenxu Wang , Jie Zhang , Jingyang Wang

{"title":"Microstructure evolution in titanium carbide with different stoichiometry under 3 MeV Au2+ ion irradiation","authors":"Jinyu Shi , Yiming Lei , Chenxu Wang , Jie Zhang , Jingyang Wang","doi":"10.1016/j.jnucmat.2025.155609","DOIUrl":"10.1016/j.jnucmat.2025.155609","url":null,"abstract":"<div><div>Titanium carbide (TiC) with the merits of stability and corrosion resistance has been regarded as promising structural material candidate for advanced nuclear reactors. The effects of deviation in carbon stoichiometry and local ordering of carbon vacancies on the irradiation-induced microstructure evolution of TiC<sub>x</sub> (<em>x</em> = 0.62–0.98) were targeted. 3 MeV Au<sup>2+</sup> ion irradiation at room temperature (RT) was conducted over a series of ion fluences ranging from 1 × 10<sup>14</sup> to 2 × 10<sup>16</sup> ions cm<sup>-2</sup>, together with grazing incidence X-ray diffraction (GIXRD) and transmission electron microscopy (TEM). No amorphization was traced for titanium carbide ceramics with different stoichiometry irradiated at doses up to ∼70 displacements per atom (dpa). Substoichiometric titanium carbides exhibited excellent lattice expansion resistance compared to near stoichiometric one beyond a dose of ∼30 dpa. In addition, irradiation-induced two ordered phases and twins were observed. Local ordering of C vacancies benefits the accommodation, annihilation of irradiation induced defects, which enhances the tolerance of irradiation-induced amorphization of titanium carbide ceramics. This work provides a comprehensive understanding of microstructure evolution in titanium carbide with different stoichiometry, which facilitates the application of titanium carbide ceramics as advanced reactors cores concepts.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"606 ","pages":"Article 155609"},"PeriodicalIF":2.8,"publicationDate":"2025-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143154924","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}

引用次数: 0