Nuclear Engineering and Design最新文献

{"title":"Study on the photoacoustic method for accurate, online, in-situ, non-contact measurement of corrosion-related unidentified deposit thickness","authors":"Guangchao Yang ,&nbsp;Xiaojing Liu ,&nbsp;Decao Wu ,&nbsp;Jinbiao Xiong ,&nbsp;Tengfei Zhang ,&nbsp;Xiang Chai ,&nbsp;Hui He","doi":"10.1016/j.nucengdes.2025.114492","DOIUrl":"10.1016/j.nucengdes.2025.114492","url":null,"abstract":"<div><div>The thickness of Corrosion-Related Unidentified Deposits (CRUD) on the surface of Pressurized Water Reactor (PWR) fuel cladding is one of the key factors affecting fuel performance. Accurate measurement of CRUD thickness is crucial for ensuring the safe operation of nuclear reactors. This paper presents an online, in-situ, non-contact CRUD thickness measurement method based on the photoacoustic effect, overcoming the issues of surface morphology changes and re-dissolution associated with traditional offline measurement methods. To validate the feasibility and accuracy of this method, an experimental system was established, and the signal transmission mechanisms involved in the thermal-fluid–solid multi-physics fields of the measurement method were explored using a simplified 1D mathematical model and 3D simulations. The results show that the method enables low-cost, high-precision measurements, requiring only a 532 nm wavelength, 700 Hz modulation frequency, and 3 W full power Pulse Width Modulation (PWM) pulsed laser as the signal excitation source, along with a piezoelectric transducer as the signal receiver. The simplified 1D mathematical model and 3D simulation results align well with experimental data, with a maximum error of ±10 %. The results show that under the same laser operating conditions, the CRUD thickness is approximately proportional to the output signal intensity with a good linear relationship. This measurement method and its mathematical model provide significant technical and theoretical support for studying the dynamic growth behavior of CRUD.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114492"},"PeriodicalIF":2.1,"publicationDate":"2025-09-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145155940","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":"Effects of grain size and porosity on cladding failure in high-burnup UO2: A sensitivity and uncertainty study","authors":"Ian Greenquist,&nbsp;Ashley Shields","doi":"10.1016/j.nucengdes.2025.114484","DOIUrl":"10.1016/j.nucengdes.2025.114484","url":null,"abstract":"<div><div>Isotopic taggants are being studied to aid in the provenance assessment of nuclear materials. However, these taggants must be selected such that they do not adversely affect fuel performance during normal operation or accident scenarios. Taggants are known to affect the fuel’s grain size and porosity. In the work described in this paper, the BISON fuel performance code was used to assess the potential effects of taggants (i.e., grain size and porosity) on fuel rod behavior and cladding failure during a high-burnup, large-break loss-of-coolant accident. Here, 281 individual fuel rods from the same reactor core were modeled for a sensitivity study, a parametric study, and uncertainty quantification.</div><div>The cladding failure predictions often exhibited stochastic behavior. After additional study, it was found that the cladding failure model is highly sensitive to residual error inherent to numerical approximation solvers. Some strategies to mitigate this sensitivity are discussed.</div><div>The study found no relationship between known taggant effects and cladding failure status. However, taggants were found to affect the time and location of failure in certain rods. Future work to continue investigating and validating these findings is briefly discussed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114484"},"PeriodicalIF":2.1,"publicationDate":"2025-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145119087","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":"Nexus between geometry and thermal-hydraulic scaling of horizontal steam generator","authors":"Milos A. Lazarevic,&nbsp;Vladimir D. Stevanovic,&nbsp;Sanja Milivojevic,&nbsp;Milan M. Petrovic","doi":"10.1016/j.nucengdes.2025.114487","DOIUrl":"10.1016/j.nucengdes.2025.114487","url":null,"abstract":"<div><div>Mature technology of large-scale nuclear power plants is a strong basis for the development of small modular and medium size reactors. Therefore, it is important to study the nexus between geometry and thermal–hydraulic scaling with the aim to take advantage of large-scale plants to develop safe and reliable scaled-down plants. The present study investigates the thermal-hydraulics of a large-scale Horizontal Steam Generator (HSG) built in the WWER 1000 type nuclear power plant, and a thermal-hydraulics of its replica with the 50 % scaled-down geometry. Unlike most previous studies, which primarily relied on the similitude concept for scaling analysis, this work employs numerical simulations with an in-house computational code based on a three-dimensional two-fluid model approach and closure laws for the prediction of interfacial transport phenomena. Obtained results show that a uniform linear reduction of HSG dimensions in three-dimensional space leads to strong non-linear changes of thermal–hydraulic parameters. Nearly the same ranges of void fraction and two-phase flow velocity changes along vertical and horizontal directions of tube bundles are obtained in the full-size HSG and in the HSG with 50% scaled-down geometry if the tube bundle volumetric heat flux in the 50% scaled-down HSG geometry is two times greater than the full-size HSG value. It is shown that a significant increase of the volumetric heat flux is achievable with a reduced diameter of tubes in the bundle and a corresponding increase of the primary side reactor coolant flow rate, although the HSG primary and secondary fluid inlet and outlet temperatures and pressure levels are kept constant. Therefore, the scaled-down HSG geometry enables significant increase of heat power per unit of tube bundle volumes, while the preserved similarity of the thermal–hydraulic conditions ensures that the scaled-down HSG operates within safe and reliable limits comparable to the full-size HSG. The findings contribute to an understanding of HSG scaling effects and support the development of small modular and medium size reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114487"},"PeriodicalIF":2.1,"publicationDate":"2025-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145119086","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 investigation of pressure drop and flow behavior in a 9 × 9 helical-cruciform fuel assembly","authors":"Ugur Karakurt ,&nbsp;Hansol Kim ,&nbsp;Joseph Seo ,&nbsp;Yassin A. Hassan","doi":"10.1016/j.nucengdes.2025.114448","DOIUrl":"10.1016/j.nucengdes.2025.114448","url":null,"abstract":"<div><div>The helical-cruciform fuel (HCF) design features helically twisted rods with a four-lobed cross-section, increasing the surface-to-volume ratio compared to conventional cylindrical rods. This geometry enhances heat transfer and coolant mixing while reducing peak fuel temperatures. The self-supporting rod arrangement eliminates spacer grids, reducing flow obstruction and pressure losses. These characteristics enable higher reactor power density and lower operating temperatures in current and next-generation reactors. This study numerically investigates pressure drop and flow characteristics in a 9 × 9 HCF assembly using Reynolds-Averaged Navier-Stokes (RANS) simulations with the <span><math><mrow><mi>k</mi></mrow></math></span>-<span><math><mrow><mi>ω</mi></mrow></math></span> Shear Stress Transport (SST) turbulence model. Preliminary analyses of 3 × 3 to 11 × 11 assemblies demonstrated that the 9 × 9 configuration is the minimum bundle size required to accurately represent essential flow behavior in larger assemblies. Entrance length analysis at Re = 21,121 shows fully developed flow after one helical pitch. The CFD model is validated against experimental pressure drop data. Flow regime boundaries are estimated at Reynolds numbers of approximately 861 (laminar-to-transitional) and 9,755 (transitional-to-turbulent). A friction factor correlation covering Re = 119 to 21,958 is developed and compared with existing correlations. The flow characteristics at Re = 21,121 were analyzed within subchannels and inter-subchannel gaps.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114448"},"PeriodicalIF":2.1,"publicationDate":"2025-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145155939","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-based evaluation of safety margins and operational limits for a boron-free small modular reactor core","authors":"Bright Madinka Mweetwa ,&nbsp;Marat Margulis","doi":"10.1016/j.nucengdes.2025.114485","DOIUrl":"10.1016/j.nucengdes.2025.114485","url":null,"abstract":"<div><div>Boron-free small modular reactor cores (SMR) experience high thermal feedback due to their high negative moderator temperature coefficients (MTC). In this work, operational limits and safety margins are proposed for a 1,429.51 MWth boron-free SMR. The core (at full power) has a span of high negative MTC values throughout the cycle with BOC, 8.5 MWd/kg, and EOC values of –32.41 pcm/K, –32.19 pcm/K, and −37.81 pcm/K respectively. A core with a high negative MTC is prone to instantaneous reactivity insertion in an event of a drop in coolant inlet temperature. In this work, a reactivity insertion accident (RIA) defined by a rupture in the main steam line has been applied as an initiating event for a drop in coolant inlet temperature. Seven cases were simulated − five cases represented a drop in coolant inlet temperature over the range of 10–50 K, and two cases represented the normal operating condition and the drop in coolant inlet temperature of 21.67 K which was the threshold for violating the heat flux hot channel factor (<span><math><mrow><msub><mi>F</mi><mi>q</mi></msub><mrow><mo>)</mo></mrow></mrow></math></span> design limit. Parameters considered were heat flux hot channel factor (<span><math><msub><mi>F</mi><mi>q</mi></msub></math></span>), core thermal power, enthalpy, fuel temperature, fuel centerline temperature, and minimum departure from nucleate boiling ratio (MDNBR). SIMULATE3-K, a thermal–hydraulic code, and SIMULATE3, a neutronics code, were coupled and used to simulate the transient. The transient was performed at end of cycle (EOC) hot full power (HFP) condition, as this condition provided the most limiting parameter response to the transient. The Rohsenow-Griffith-Kutateladze (RGK) correlation was employed to calculate the critical heat flux (CHF) applicable to the fuel pin heat transfer regime. A combination of best estimate and conservative estimates was applied in establishing operational limits and safety margins. Nominal values for the <span><math><msub><mi>F</mi><mi>q</mi></msub></math></span>, fuel temperature, fuel centerline temperature, enthalpy and DNBR were found to be consistent and within the range of conventional light water reactor operational limits and safety margins. The most limiting parameter was observed to be the <span><math><msub><mi>F</mi><mi>q</mi></msub></math></span>, whose design limit of 2.6 was violated with a drop in coolant inlet temperature of 21.67 K. A conservative drop in coolant inlet temperature of 10 K was proposed as an operational limit for the boron-free SMR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114485"},"PeriodicalIF":2.1,"publicationDate":"2025-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145155931","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":"Thermal mechanical assessment of a SiC-SiC-composite clad fuel pin concept in a light water reactor environment","authors":"John Alvis ,&nbsp;Gyanender Singh ,&nbsp;Kyle Gamble ,&nbsp;David Kamerman ,&nbsp;Seokbin Seo ,&nbsp;Katheren Nantes","doi":"10.1016/j.nucengdes.2025.114458","DOIUrl":"10.1016/j.nucengdes.2025.114458","url":null,"abstract":"<div><div>Accident-tolerant fuels (ATFs) are designed to increase coping time following an accident scenario while preserving or improving current steady-state reactor operational performance. A potential ATF concept is using silicon carbide (SiC)-SiC-composite claddings. Fuel-performance simulations were conducted on a SiC-SiC-based cladding concept utilizing a multilayered approach for improved performance. This cladding concept is referred to in this paper as “the duplex concept” as it is a duplex structure composed of a monolithic SiC layer placed on the outside of an inner SiC-SiC composite layer. The monolithic SiC layer is used to provide gas tightness to the rod and protect the SiC-SiC-composite layer from exposure to the coolant. A liquid metal is added to the fuel-cladding gap for improved thermal transport between the fuel and the cladding. In this work, the BISON fuel-performance code was used to conduct fuel-performance simulations on the cladding concept. Comparisons are made with a current prototypic fuel-rod design consisting of uranium dioxide (UO₂) fuel enclosed in Zircaloy-4 cladding under four relevant conditions. For condition I (normal operations) two representative steady-state cases were considered, one with a constant rod average heat rate, and one with an initially higher heat rate. For condition II events, a pellet-cladding interaction (PCI) ramp case was simulated to analyze potential anticipated operational occurrences. Condition III/IV transient responses during a loss of coolant accident (LOCA) and a reactivity-initiated accident (RIA) were also simulated. This computational study demonstrated that for normal operating conditions, the SiC concept cladding performed as well as the baseline for the standard-power cases evaluated. The ramping evaluations indicate potential for earlier fracturing of the SiC-SiC composite cladding compared to the Zircaloy-4 cladding due to the temperature gradient and the subsequent differential thermal conductivity degradation and swelling across the composite thickness. In condition III/IV events the SiC-SiC duplex concept remains intact after a 700 J/g RIA similar to Zircaloy-4 fuel systems. Under LOCA conditions, the duplex concept showed significantly improved performance, remaining intact in contrast to Zircaloy-4 which ballons and bursts.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114458"},"PeriodicalIF":2.1,"publicationDate":"2025-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145156045","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":"Development of a thermal creep model for aluminum alloy 6061 cladding in U-10Mo monolithic fuel plates","authors":"Revanth Mattey ,&nbsp;Alexander Swearingen ,&nbsp;Hakan Ozaltun ,&nbsp;Jeffrey J. Giglio","doi":"10.1016/j.nucengdes.2025.114428","DOIUrl":"10.1016/j.nucengdes.2025.114428","url":null,"abstract":"<div><div>Plate-type fuel elements consisting of a high-density, low-enriched uranium (LEU) U–10Mo-based fuel foil encapsulated in an aluminum alloy (AA) cladding are fabricated using the hot isostatic pressing (HIP) technique. During the HIP process, the fuel plate system is heated to 560 °C, then cooled to room temperature. This heat cycle significantly affects the mechanical properties of the aluminum cladding, and experimental investigations have shown that, post-HIP bonding, the mechanical properties of the aluminum cladding transition from those of AA 6061-T6 to something closer to the O temper. More specifically, the ultimate strength of the cladding decreases while its ductility increases, making it challenging to capture the changes in mechanical behavior and material properties. Understanding the residual stresses generated during the HIP process is critical for assessing the fuel plate’s integrity under various temperature, pressure, and irradiation. To simulate the HIP bonding process, the elastic, plastic, and thermal properties of the cladding are assumed to be similar to those of AA 6061-O temper. However, the primary challenge lies in the lack of available data for the creep model of the AA 6061 cladding during this transient process of HIP. The present study focuses on developing a computational model that predicts the creep behavior of the aluminum cladding in the fuel plates during the HIP process, as cladding creep significantly influences the residual stresses generated in U-10Mo fuel plates during HIP fabrication. Furthermore, as HIP bonding occurs at high temperatures that are nearing the melting point of aluminum, the present work considered a temperature-dependent Arrhenius-type creep model. In particular, a hyperbolic sine creep model is employed to estimate the creep properties of the as-fabricated aluminum cladding. The residual stresses predicted in the U-10Mo fuel when using the newly calibrated creep model closely align with the experimental measurements, validating the model’s accuracy.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114428"},"PeriodicalIF":2.1,"publicationDate":"2025-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145119085","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":"An empirical method for inverse uncertainty quantification in nuclear thermal–hydraulic codes","authors":"Andrea Bersano ,&nbsp;Francesco Di Maio ,&nbsp;Enrico Zio ,&nbsp;Nicola Pedroni ,&nbsp;Fulvio Mascari","doi":"10.1016/j.nucengdes.2025.114483","DOIUrl":"10.1016/j.nucengdes.2025.114483","url":null,"abstract":"<div><div>The quantification of the uncertainty in the output of the thermal–hydraulic and severe accident codes used for the safety analysis of Nuclear Power Plants (NPPs) typically proceeds by propagating the uncertainty of the input variables and parameters through the codes. The input uncertainty is usually characterized by expert-based probability density functions. Inverse Uncertainty Quantification (IUQ) can be used to inform the characterization of such probability density functions based on experimental data. In this paper, we propose an empirical IUQ method that exploits data from experimental facilities and apply it to a case study of the ATRIUM project of OECD/NEA/CSNI/WGAMA. Specifically, the TRACE best-estimate thermal–hydraulic system code is used to replicate the data of the Sozzi-Sutherland, Super Moby Dick and Marviken experimental facilities with regards to an accident scenario with guillotine break of the pressurizer surge line in a generic three loop PWR-900; the input uncertainties are first characterized and then propagated to the output.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114483"},"PeriodicalIF":2.1,"publicationDate":"2025-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145119083","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":"Research on the IWF characteristics of LFR under normal operation and blockage conditions","authors":"Di Wu ,&nbsp;Minyang Gui ,&nbsp;Jie Cheng ,&nbsp;Jianjun Wang","doi":"10.1016/j.nucengdes.2025.114481","DOIUrl":"10.1016/j.nucengdes.2025.114481","url":null,"abstract":"<div><div>The prerequisite for the engineering application of lead-based fast reactors (LFR) is to ensure the safe and reliable operation of the reactor; thus, it is crucial to conduct research on the thermal safety characteristics of LFR. In this study, the inter-wrapper flow (IWF) characteristics of LFR under normal operation and blockage conditions were analyzed with the developed sub-channel code. The developed sub-channel code was verified with the existing experimental data, and the deviation is within ± 10°C, which means the accuracy of the code is acceptable. Furthermore, the influence of IWF, assembly flow rate, and the power of the whole rod bundle on the thermal–hydraulic parameters was studied. The impact of changes in the thermal parameters of individual assemblies was considered as well, including the condition where only one of three rod bundles’ flow rate is 50 % of the original, and the whole heating power is 20 % of the original. Finally, the local flow blockage accidents were analyzed, and the different blockage sub-channels and blockage areas were investigated. The research could provide a reference for the thermal analysis and optimization design of LFR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114481"},"PeriodicalIF":2.1,"publicationDate":"2025-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145119082","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":"Radiation damage simulation of reactor pressure vessel material: A multi-scale approach","authors":"Genshen Chu ,&nbsp;Dandan Chen ,&nbsp;Kang Zou ,&nbsp;Xu Lu","doi":"10.1016/j.nucengdes.2025.114415","DOIUrl":"10.1016/j.nucengdes.2025.114415","url":null,"abstract":"<div><div>RPV (reactor pressure vessel) is the only core equipment that cannot be replaced during the life of the reactor. Radiation embrittlement is its main performance issue, and the precipitation of Cu-rich clusters is one of the main reasons for the embrittlement of RPV steel. In this paper, the multi-scale model is established and its simulation software MISA is developed for simulating evolution from primary defect generation to long-term evolution of defects and macro embrittlement impact, which can provide effective reference for the research and development of reactor materials.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114415"},"PeriodicalIF":2.1,"publicationDate":"2025-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145119084","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}