Progress in Nuclear Energy最新文献

{"title":"Study on enhancement methods for fault feature detection in rotating machinery within nuclear power plant","authors":"Hang Wang ,&nbsp;Hailin Song ,&nbsp;Fangxiaozhi Yu ,&nbsp;Xinxing Wang ,&nbsp;Zhao Xu","doi":"10.1016/j.pnucene.2026.106275","DOIUrl":"10.1016/j.pnucene.2026.106275","url":null,"abstract":"<div><div>Pump equipment constitutes a critical element within a nuclear power plant, with its reliability being essential for continuous and stable operation of facility. Throughout the extended operational lifespan in nuclear power plants, the key rotating components of pump equipment are susceptible to failures, including cracks and wear, which can impair overall performance. The operational environment of pump equipment is characterized by significant levels of noise, which leads to fault characteristics being very weak in collected signals. However, existing methods for enhancing fault characteristics, such as Local Mean Decomposition (LMD), are unable to eliminate noise interference in signals. This article proposes an adaptive endpoint extension LMD method aimed at addressing several challenges associated with existing algorithms. These challenges include the difficulty in distinguishing components of fault information, poor enhancement of fault features, and low robustness of the model. Additionally, to further improve fault characteristics, an adaptive iterative weighting model for fault characteristic enhancement is developed based on the principles of stochastic resonance (SR). Finally, a comparative analysis of experimental data is conducted to demonstrate superiority of the model.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"195 ","pages":"Article 106275"},"PeriodicalIF":3.2,"publicationDate":"2026-05-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146161683","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Transient and mechanical analysis of phenomena of supercritical carbon dioxide jet into lead-bismuth eutectic in lead-cooled fast reactor","authors":"Haoyang Li ,&nbsp;Dalin Zhang ,&nbsp;Xinyu Li ,&nbsp;Boyang Jiang ,&nbsp;Xiaoyan Yang ,&nbsp;Pengrui Qiao ,&nbsp;Wenxi Tian ,&nbsp;Suizheng Qiu ,&nbsp;G.H. Su","doi":"10.1016/j.pnucene.2026.106297","DOIUrl":"10.1016/j.pnucene.2026.106297","url":null,"abstract":"<div><div>The rupture of heat exchangers in Lead-cooled Fast Reactors (LFRs) can lead to the injection of high-pressure supercritical carbon dioxide (sCO₂) into liquid lead-bismuth eutectic (LBE), constituting a severe accident scenario that threatens reactor integrity. Given the limited research on this interaction, this study performs a systematic numerical investigation of the transient thermal-hydraulic behavior and structural response within a designed experimental vessel. Using a Mixture Multiphase (MMP) CFD model with real-gas properties, the injection of a 16 MPa and 450 °C sCO₂ jet into an LBE pool was analyzed. Results indicate that high-velocity shear flows induce intense viscous heating, while the pressure field exhibits non-periodic decay driven by viscous dissipation and acoustic radiation. To evaluate structural safety, the transient pressure history was coupled with an acoustic-structural analysis model. The assessment reveals that while the ductile stainless steel vessel remains within elastic limits, the brittle sapphire window experiences tensile stresses exceeding its failure threshold, particularly at the bi-material interface. These findings identify critical failure mechanisms in brittle observation components and provide essential theoretical guidance for the safety optimization of sCO₂–LBE systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"195 ","pages":"Article 106297"},"PeriodicalIF":3.2,"publicationDate":"2026-05-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146161681","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Atmospheric dispersion and ecological risk assessment of hypothetical accidents at Kenya's first proposed nuclear power plant","authors":"Osamong Gideon Akou ,&nbsp;Xuan Wang ,&nbsp;Shuhuan Liu ,&nbsp;Xinwei Liu ,&nbsp;Guanghui Su ,&nbsp;Ailing Zhang ,&nbsp;Junfang Zhang ,&nbsp;Minghua Lv ,&nbsp;Lei Huang ,&nbsp;Shanchao Yang","doi":"10.1016/j.pnucene.2026.106294","DOIUrl":"10.1016/j.pnucene.2026.106294","url":null,"abstract":"<div><div>Kenya's nuclear energy ambitions require rigorous safety and environmental assessments for proposed sites such as the Kilifi Nuclear Power Plant (NPP). This study models radiological consequences of a severe station blackout-induced LOCA using RASCAL (source term), HotSpot (atmospheric dispersion), and ERICA (ecological risk). Kilifi's coastal climate presents neutral stability (mean wind speed: 4.86 m/s) but complex dispersion behavior. The release was dominated by noble gases (75.6%) and iodine (24.2%). Peak Total Effective Dose (TED) reached 4.3 × 10⁻³ Sv under Class C conditions, with mitigation reducing TED by ∼99.9%. Terrain shielding reduced I-131 groundshine by over tenfold in urban areas. Raising release height from 50 m to 100 m reduced TED by 99%, with limited benefit above 200 m. Ecological risks were highest for Cs-137 (aquatic, RQ 66.9%) and Ru-106 (sediment, RQ 68.9%). HotSpot validation using Kincaid SF₆ data confirmed its conservative reliability. Findings inform Kenya's nuclear siting, emphasizing iodine protection, terrain-aware planning, and long-term environmental monitoring.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"195 ","pages":"Article 106294"},"PeriodicalIF":3.2,"publicationDate":"2026-05-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146161684","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"A decision-support framework for nuclear facility siting using machine learning and half quadratic programming: Insights from Türkiye","authors":"Ayhan Doğan","doi":"10.1016/j.pnucene.2026.106296","DOIUrl":"10.1016/j.pnucene.2026.106296","url":null,"abstract":"<div><div>This study introduces an innovative methodological framework for nuclear power plant (NPP) site selection in Mersin Province, integrating the three best-performing Machine Learning (ML) algorithms from a set of seven, each offering distinct advantages, with Half Quadratic Programming (HQP) and Geographic Information Systems (GIS). The study achieved the most favorable outcomes with Categorical Boosting (CB), Support Vector Machine (SVM), and Multi-Layer Perceptron (MLP). The integration of these algorithms, SVM (R² = 0.960), MLP (R² = 0.953), and CB (R² = 0.912), significantly enhanced the accuracy and objectivity in determining criterion weights, thereby reducing reliance on subjective expert opinions. HQP effectively harmonized the weights derived from different ML methods, optimizing critical criterion interactions such as seismic risk, environmental factors, and infrastructure efficiency. GIS enabled precise spatial analyses, identifying coastal areas around Anamur and Aydıncık as most suitable (score 5), and inland regions near Gülnar and Silifke as highly suitable (score 4). Notably, the existing location of the Akkuyu NPP precisely matches the area identified as most suitable, validating the robustness and reliability of the proposed approach. Overall, this research significantly advances nuclear site selection methodologies, offering a comprehensive, objective, and practical decision-making tool that substantially contributes to strategic energy planning.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"195 ","pages":"Article 106296"},"PeriodicalIF":3.2,"publicationDate":"2026-05-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146161682","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Numerical simulation study on the coupled heat transfer characteristics of liquid sodium and supercritical carbon dioxide","authors":"Haonan Zhang ,&nbsp;Zhengyu Gong ,&nbsp;Songbai Cheng ,&nbsp;Yanan Zhao","doi":"10.1016/j.pnucene.2026.106293","DOIUrl":"10.1016/j.pnucene.2026.106293","url":null,"abstract":"<div><div>Among fourth-generation nuclear technology, the Sodium-cooled Fast Reactor (SFR) is one of the most mature reactors, offering high thermal efficiency and inherent safety advantages. The adoption of supercritical carbon dioxide (S-CO₂) as the secondary-loop heat transfer medium in SFR has gained significant attention due to its potential to reduce heat exchanger size and improve thermal-to-electric conversion efficiency. Furthermore, the coupled system of an S-CO₂ Brayton cycle with an SFR effectively eliminates sodium-water reaction risks while achieving higher cycle efficiency compared to the steam Rankine cycle. Meanwhile, the Printed Circuit Heat Exchanger (PCHE), known for its compact structure and superior heat transfer performance, is considered as a promising candidate for thermal coupling between sodium and S-CO₂. However, studies on heat transfer between sodium and S-CO₂ in PCHE remain scarce, and the associated technologies are still immature, necessitating further investigation into their thermal interaction and fundamental design guidelines. This study numerically investigates the heat transfer characteristics of liquid sodium and S-CO₂ through geometric modeling and simulation of a PCHE. Parameter-controlled simulations were performed to assess the impact of flow conditions and channel geometry on temperature and pressure distributions. The results indicate that temperature fields are primarily influenced by the inlet mass flow rates and temperatures of both fluids, while S-CO₂ outlet pressure, channel cross-section, and horizontal spacing show negligible effects (the variation in outlet temperature remains below 8 K). In contrast, pressure drop is mainly affected by the inlet mass flow rate of each fluid and the channel cross-sectional area, with an increase in channel diameter from 1.6 mm to 2.4 mm reducing the pressure drop by over 80%. The heat transfer coefficient is highly sensitive to changes in S-CO₂ mass flow rate and channel cross-sectional area; within the studied ranges, it increased by 58% and decreased by 44% respectively, both leading to significant performance variation. This study provides valuable insights for the design of sodium-S-CO₂ heat exchangers and supports the development of SFR technology.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"195 ","pages":"Article 106293"},"PeriodicalIF":3.2,"publicationDate":"2026-05-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146162163","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental study and software simulation of air-water periodic flooding in a 3×3 bundle channel","authors":"Guangyuan Jin ,&nbsp;Weilian Li ,&nbsp;Jinghu Bai ,&nbsp;Rui Wang ,&nbsp;Yandong Hou","doi":"10.1016/j.pnucene.2026.106262","DOIUrl":"10.1016/j.pnucene.2026.106262","url":null,"abstract":"<div><div>Investigating the behavior of gas-liquid two-phase flow within rod bundle channels can provide theoretical support for the smooth operation and emergency response of reactor systems. The characteristics of flow in a 3 × 3 rod bundle channel under periodic flooding conditions were investigated using numerical simulation methods. Comparing the simulation results with the experimental data revealed a high degree of agreement in the gas-liquid behavior and the time fractions across different flooding regions. This study examines the evolution of the void fraction along the axial direction and the variation of liquid phase velocity within the subchannels. The study of the local liquid phase velocity field analyzes the mechanisms underlying the onset of the flooding front and the formation of liquid phase accumulation. The relationship between the dimensionless pressure gradient and liquid holdup was analyzed, and an examination of the force conditions on the liquid film led to the conclusion that periodic flooding results from the alternating effects of the gas phase's carrying force and the gravity of the liquid phase in different regions. These research findings further enhance the understanding of thermal-hydraulic characteristics within rod bundle channels.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106262"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145963369","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Review of the development of turbulence models for low Prandtl number fluid flows and heat transfer based on machine learning techniques","authors":"Ikram Ul Haq,&nbsp;Mohsin Raza,&nbsp;Waqar Ul Hassan,&nbsp;Li-Qi Guo,&nbsp;Jun-Liang Guo,&nbsp;Hong-Na Zhang,&nbsp;Xiao-Bin Li,&nbsp;Feng-Chen Li","doi":"10.1016/j.pnucene.2026.106264","DOIUrl":"10.1016/j.pnucene.2026.106264","url":null,"abstract":"<div><div>Nuclear energy is a clean and sustainable power source with minimal environmental impact. Its effectiveness relies on advanced cooling systems that use low Prandtl number (<span><math><mrow><mi>P</mi><mi>r</mi></mrow></math></span>) fluids, such as liquid metals, to manage extreme heat in liquid-metal cooled fast reactors (LMFRs). However, predicting the turbulent heat transfer and fluid flow behaviour of these fluids is challenging for traditional Reynolds-averaged Navier–Stokes (RANS) models due to their non-equilibrium turbulence and complex flow patterns. While large-eddy simulation (LES) and direct numerical simulation (DNS) methods are sufficiently accurate, they remain impractical for engineering applications due to their high computational cost. Machine learning (ML) offers innovative solutions to improve turbulence modelling, enabling more accurate predictions and efficient designs. This review highlights the need to advance ML-assisted turbulence modelling for low-<span><math><mrow><mi>P</mi><mi>r</mi></mrow></math></span> fluids by reviewing traditional models, ML-augmented approaches and their real-world applications. It also addresses current limitations such as limited data, model reliability and generalization issues, while outlining opportunities to develop smarter, safer and more efficient nuclear thermal hydraulics cooling systems for the nuclear energy sector and other cutting-edge technologies.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106264"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145963368","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Upgrade of FROBA code to derive new insights in fuel performance of Lead-Bismuth Eutectic Cooled Fast Reactor","authors":"Yangbin Deng ,&nbsp;Zijian Huang ,&nbsp;Cong Liu ,&nbsp;Ruikuan Li ,&nbsp;Yuefeng Wu ,&nbsp;Yingwei Wu ,&nbsp;Guanghui Su","doi":"10.1016/j.pnucene.2026.106252","DOIUrl":"10.1016/j.pnucene.2026.106252","url":null,"abstract":"<div><div>Lead-Bismuth Eutectic Cooled Fast Reactor (LBEFR) uses low melting point and high boiling point lead-bismuth as the coolant, and is one of the most promising Generation IV nuclear systems. However, fuel rods in LBEFR have to be subjected to more severe operation conditions compared with Light Water Reactor (LWR) fuels, such as harder neutron spectrum, higher temperature, and enhanced coolant corrosiveness, resulting in distinct evolution processes of fuel behavior. In this study, with full consideration of geometry and material design characteristics, models on distinct phenomena of LBEFR fuels, such as fuel constituent migration, pellet restructuring, cladding corrosion, have been developed and implanted into the program FROBA initially developed for LWR fuel analysis. The capability of the upgraded FROBA for LBEFR fuel performance analysis was preliminarily verified with several benchmark cases. Subsequently, a fuel behavior simulation was conducted for MOX-T91 fuel in a compact LBEFR under long-term normal operation and a power ramp following the normal operation. Compared with traditional LWR fuels, the LBEFR fuel operated at significantly higher temperatures, triggering substantially more fission gas release. To accommodate the large volume of released fission gases and suppress internal pressure rise, LBEFR fuel requires a large plenum design. Benefiting from the coolant outlet temperature controlled around 500 °C, the corrosion of T91 cladding was limited, resulting in a minimal oxide layer thickness that poses no significant threat to cladding structural integrity. However, at high burnup stages, the T91 cladding exhibited pronounced irradiation swelling specifically within the critical temperature range of 420 ± 40 °C. To ensure cladding deformation remains within acceptable limits, T91 should avoid operation within this swelling-sensitive temperature range. Power ramp sensitivity analysis reveals that elevated fuel temperatures drive fuel restructuring-induced central void formation, significantly influencing the thermo-mechanical response. In specific, the central void formation could reduce fuel center temperatures and alleviate contact pressure during pellet-cladding mechanical interaction. These findings provide valuable insights crucial for the design and optimization of LBEFR fuels.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106252"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981519","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Effect of channel geometry on the onset of significant void and void fraction profiles of vertical upward boiling flows","authors":"Shichang Dong,&nbsp;Takashi Hibiki","doi":"10.1016/j.pnucene.2026.106256","DOIUrl":"10.1016/j.pnucene.2026.106256","url":null,"abstract":"<div><div>The assessment of heat transfer performance characteristics relies on the accurate prediction of void fraction in heat transfer systems. Existing typical void fraction models rely on the determination of the point of onset of significant void (OSV), which limits their ability to estimate the development of void fraction before the OSV point. To address this limitation, the present study first developed a flow quality model for the subcooled boiling region. This model was then extended to the saturated boiling region, resulting in a flow quality model applicable across the entire boiling region. Subsequently, this model was integrated with a drift-flux correlation (DFC) to formulate a predictive model for axial void fraction profiles in vertical boiling flows in annular channels. The validation of the developed model was conducted using a comprehensive experimental database including 1300 data points from five independent sources with water as the working fluid. Results showed that, compared to the existing typical model, the developed model could successfully predict the evolution of the axial void fraction upstream of the OSV point. This model also demonstrated superior accuracy in predicting the OSV point and downstream void fraction behavior, thereby significantly enhancing the prediction accuracy of the axial void fraction profiles. Furthermore, it accurately predicted the influence of key thermal-hydraulic parameters on the void fraction profiles. This paper also revealed the impact of channel geometry by comparing OSV and void fraction profiles across circular, rectangular, and annular channels.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106256"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039087","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Numerical study on heat transfer enhancement of LBE flow in semicircular-fin fuel bundles","authors":"Qian Li ,&nbsp;Siwei Cai ,&nbsp;Shengcai Zhang ,&nbsp;Xuechen Liu ,&nbsp;Nianmei Zhang ,&nbsp;Chen Hu ,&nbsp;Xian Zeng ,&nbsp;Jianchuang Sun ,&nbsp;Weihua Cai","doi":"10.1016/j.pnucene.2026.106253","DOIUrl":"10.1016/j.pnucene.2026.106253","url":null,"abstract":"<div><div>In this study, numerical simulation methods were employed to investigate the flow and heat transfer characteristics of liquid lead-bismuth eutectic (LBE) alloy in novel fuel assemblies with different fin winding directions of the fuel rod bundle. The axial and circumferential thermal-hydraulic parameter data of subchannels and fuel rods were extracted through micro-segment methods and sub-channel partitioning. The research results showed that changing the fin winding direction had a minor effect on the pressure drop of the fuel assembly, while the heat transfer coefficient increased by 15 % compared to the original fuel assemblies, demonstrating that the new fuel assemblies effectively enhance the heat transfer capacity of LBE in semicircular-fin rod bundles. Further investigation into the mechanisms of enhanced heat transfer in novel fuel assemblies reveals that altering the orientation of the fins results in co-directional flow of the LBE. This co-directional flow can increase the secondary flow velocity while maintaining the <em>Q</em> invariant as a positive value, thereby reducing the occurrence of rotational flow and enhancing heat transfer capabilities. This study offers new research perspectives on the design of lead-bismuth fast reactor fuel assemblies and the analysis of their thermal-hydraulic characteristics, which are of significant importance for the structural design of lead-bismuth fast reactors and for the mechanistic research of nuclear reactors.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106253"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039148","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}