Journal of Nuclear Materials最新文献

{"title":"Microstructure, oxidation kinetics and hydrogen absorption of Cr-coated Zr-Sn-Nb alloy cladding tubes after single-sided oxidation at 1000–1200 °C followed by fast reflood","authors":"Weiwei Xiao ,&nbsp;Sheng Xu ,&nbsp;Xiao Hu ,&nbsp;Jinghao Huang ,&nbsp;Shihong Liu ,&nbsp;Shuliang Zou","doi":"10.1016/j.jnucmat.2025.155718","DOIUrl":"10.1016/j.jnucmat.2025.155718","url":null,"abstract":"<div><div>Reflood of nuclear fuel assemblies is the top priority accident management strategy for nuclear power plants in the event of a loss of coolant accident, during which the cladding tubes inevitably undergo reflood oxidation. This study aims to investigate the single-sided reflood oxidation behavior of Cr-coated Zr-Sn-Nb alloy cladding tubes at 1000 °C-1200 °C. High-temperature steam oxidation and in-situ quenching were employed to simulate the reflood oxidation process of nuclear fuel assembly cladding tubes in the early stages of severe accidents. The microstructure, cross-sectional layer thickness evolution, oxidation kinetics, and hydrogen absorption of Cr-coated Zr-Sn-Nb alloy cladding tubes during single-sided reflood oxidation process were investigated. The results showed that after single-sided reflood oxidation, microcracks appeared on the surface of the cladding tubes. As the oxidation temperature increases and the oxidation time prolongs, the surface oxidation products gradually evolve from porous flocculent structures to strip-shaped or elliptical bubble structures and worm aggregated structures. A multi-layer layered structure of Cr₂O₃ layer/Cr coating/Cr-Zr diffusion layer/α-Zr(O) was formed on the cross-section of the cladding tube after single-sided reflood oxidation. The thickness of the Cr₂O₃ layer and residual Cr coating does not increase or decrease monotonically with the extension of oxidation time after reflood oxidation at 1200 °C. The kinetics of single-sided reflood oxidation follows a parabolic law, and the oxidation constant increases by about an order of magnitude as the oxidation temperature increases by 100 °C. As the oxidation temperature increases and oxidation time prolongs, the hydrogen absorption of the cladding tube gradually increases. After single-sided reflood oxidation, the hydrides in the Zr-Sn-Nb alloy cladding tube are mainly δ-ZrH_1.5.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155718"},"PeriodicalIF":2.8,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143535341","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":"Cesium and iodine speciation in irradiated UO2 fuel","authors":"J.-Y. Colle ,&nbsp;J.N. Zappey ,&nbsp;O. Beneš ,&nbsp;M. Cologna ,&nbsp;T. Wiss ,&nbsp;R.J.M. Konings","doi":"10.1016/j.jnucmat.2025.155715","DOIUrl":"10.1016/j.jnucmat.2025.155715","url":null,"abstract":"<div><div>The presence of CsI in nuclear fuel has long been debated. Its formation significantly decreases volatility, thereby reducing the rate at which iodine and cesium are released from the reactor core during a nuclear accident. A series of samples were investigated by Knudsen Effusion Mass Spectrometry (KEMS) in order to determine whether CsI is present in irradiated nuclear fuel. The examined samples were pure CsI, CsI exposed to gamma radiation, CsI-doped UO₂ simulated fuel and irradiated LWR fuel samples. The CsI and CsI-doped samples were examined to establish boundary conditions for the detection of CsI by KEMS. These samples indicated that the presence of CsI in fuel is characterized by three mass spectrometric signals Cs⁺, I⁺ and CsI⁺, with a peak ratio of CsI⁺ and I⁺ of 1:0.7. The examinations of irradiated fuels showed none of these characteristics and hence no evidence that CsI is present in irradiated LWR nuclear fuel, at least after a storage period of years.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155715"},"PeriodicalIF":2.8,"publicationDate":"2025-02-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143548945","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":"Fuel performance simulations of TRISO particle geometries derived from XCT","authors":"M. Poschmann,&nbsp;A. Prudil,&nbsp;R. Osmond","doi":"10.1016/j.jnucmat.2025.155714","DOIUrl":"10.1016/j.jnucmat.2025.155714","url":null,"abstract":"<div><div>The current AGR TRISO fuel specification effectively assumes that the layer thickness variations within a particle do not significantly affect particle performance. However, the limits of this assumption and their relevance for commercial TRISO production have not been established. In this work, a method was developed to generate 3D geometries of TRISO particles, including the spatial variation in layer thickness, from X-ray computed tomography for use in fuel performance modelling. Simulated irradiation of a demonstration particle found SiC hoop stress values peaking at 315 MPa in tension, significantly in excess of those from previous modelling studies with similar particle aspect ratios. Simulations with representative 2D axisymmetric geometries based on the demonstration particle predicted significantly lower stresses for the same simulated irradiation. 2D radial segments extracted with an arbitrarily oriented polar axis under-predicted the maximum SiC hoop stress by 315-400 MPa, while those extracted with the polar axis passing through the point of maximum SiC hoop stress in the 3D model under-predicted the maximum SiC hoop stress by 165-275 MPa. The 2D model produced using existing methods for generating a 2D flat-spot particle under-predicted the maximum SiC hoop stress by 215 MPa. These findings suggest that existing models may underestimate the stress caused by the asphericity of certain TRISO particle morphologies, and that the current AGR specification may not capture all of the geometric factors that contribute to particle failure probability.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155714"},"PeriodicalIF":2.8,"publicationDate":"2025-02-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143511580","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":"The effect of precipitates and alloying elements on γ-Fe (111) surface dissolution corrosion in liquid lead-bismuth eutectic by first-principles study","authors":"Yufei Li ,&nbsp;Runyu Zhou ,&nbsp;Tao Gao ,&nbsp;Changan Chen","doi":"10.1016/j.jnucmat.2025.155716","DOIUrl":"10.1016/j.jnucmat.2025.155716","url":null,"abstract":"<div><div>The M₂₃C₆ precipitates play an important role in the corrosion behavior of austenitic stainless steel. Here, by establishing the connection between the surface and the precipitated phase, this work assesses the surface dissolution corrosion by applying the electrode potential. Firstly, the influence of different carbides precipitates on surface dissolution corrosion is compared, which shows that Cr₂₂FeC₆ has the highest electrode potential (+1.68 V), accelerating surface dissolution corrosion. Then, the effect of solute atoms (Pb/Bi/O) on surface dissolution corrosion is studied. It is found that Pb/Bi will promote the dissolution of surface Fe atoms. However, O will strengthen the corrosion resistance of the surface. Simultaneously, the O inhibits the hybridization of the 3d orbital of Fe and the 6p orbital of Bi/Pb, mitigating the corrosion of Pb/Bi on the surface. Lastly, the effects of common alloying elements (Al, Si, and Ni) in austenitic steel on the corrosion of the surface are also investigated to improve surface corrosion resistance. This research attempts to provide a more comprehensive knowledge of the corrosion of iron substrates in ADSs, improving the safety of nuclear energy systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155716"},"PeriodicalIF":2.8,"publicationDate":"2025-02-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143528685","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":"Statistical fracture behavior of doped UO2 using a ball-on-ring equibiaxial flexure test method","authors":"Adrianna E. Lupercio ,&nbsp;Tashiema L. Ulrich ,&nbsp;Andrew T. Nelson ,&nbsp;Brian J. Jaques","doi":"10.1016/j.jnucmat.2025.155713","DOIUrl":"10.1016/j.jnucmat.2025.155713","url":null,"abstract":"<div><div>Metal oxide dopants, such as titanium and chromium oxides, have garnered considerable attention for their potential to increase grain size (≥ 30 µm) in UO₂ fuel, purportedly enhancing fission gas retention during reactor operation. Fuel performance is significantly impacted by fuel fracture behavior, so it is important to understand the effects of enhanced grain size and dopant content on UO₂ fuel fracture. UO₂ pellets were doped with 0.1 wt% TiO₂ and 0.3 wt% Cr₂O₃ to alter density and grain size. Inductively coupled plasma mass spectroscopy measured dopant levels pre- and post-sintering. X-ray diffraction revealed lattice changes and microstrain via Rietveld refinement. Field emission scanning electron microscopy determined grain sizes of approximately 30 µm for TiO₂ doping and 7 µm for Cr₂O₃ doping. Transverse rupture strength tests were performed on over 30 samples per dataset to obtain characteristic strength and Weibull modulus. Results indicate no statistical difference in fracture strength between 0.1 wt% TiO₂ doped UO₂ and undoped UO₂, while 0.3 wt% Cr₂O₃ doped UO₂ exhibited a 20% decrease in fracture strength. Doped UO₂ samples also showed reduced Weibull modulus compared to undoped UO₂, suggesting increased scatter in fracture strength. This study's findings suggest that titanium and chromium oxide doping in UO₂, regardless of grain size, induce residual stresses, decreasing fracture strength and increasing variability in fracture behavior.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155713"},"PeriodicalIF":2.8,"publicationDate":"2025-02-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143548941","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":"In situ self-ion (Fe+) irradiation of ODS-FeCrAl alloy fuel cladding materials with different Cr contents: The early stages of Cr-rich α’ phase precipitation","authors":"Hoang Le ,&nbsp;Yann de Carlan ,&nbsp;David T. Hoelzer ,&nbsp;Kan Sakamoto ,&nbsp;Per O.Å. Persson ,&nbsp;Jonathan A. Hinks ,&nbsp;Konstantina Lambrinou","doi":"10.1016/j.jnucmat.2025.155706","DOIUrl":"10.1016/j.jnucmat.2025.155706","url":null,"abstract":"<div><div>Oxide-dispersion-strengthened FeCrAl (ODS-FeCrAl) alloys are candidate accident-tolerant fuel cladding materials for light water reactors because they demonstrate satisfactory resistance to materials degradation effects such as high-temperature oxidation, radiation-induced swelling, and creep. Their perspective deployment to market is challenged, however, by their inherent susceptibility to irradiation embrittlement caused by the precipitation of the brittle Cr-rich α’ phase at relatively low temperatures (≤475 °C). This work used <em>in situ</em> self-ion irradiation (150 keV Fe⁺) in a transmission electron microscope to elucidate the early stages of Cr-rich α’ phase precipitation in three candidate ODS-FeCrAl alloy fuel cladding materials with different Cr contents (10, 12, and 20 wt.%) and microstructures. The early stages of the process resulting in the precipitation of the Cr-rich α’ phase in these three ODS-FeCrAl alloys under Fe⁺ irradiation were investigated at room temperature and 300 °C up to total fluences of 1.7 × 10¹⁵ ions·cm^-2 (2 dpa) and 3.4 × 10¹⁵ ions·cm^-2 (4 dpa), using three damage dose rates (5 × 10^–5, 3.3 × 10^–4, and 2 × 10^–3 dpa·s^-1). Post-irradiation examination via scanning transmission electron microscopy, energy-dispersive X-ray spectroscopy and electron energy loss spectroscopy suggested that the precipitation of the Cr-rich α’ phase might be promoted by the phase separation of the alloy matrix into Cr-rich and Fe-rich regions. Interestingly, oxygen impurities segregated preferentially in the Cr-rich regions, possibly promoting the radiation-assisted formation of the Cr-rich α’ phase. α’ phase precipitation was more pronounced at room temperature when compared to 300 °C, and it was clearly promoted by the progressive increase in the Cr content of the ODS-FeCrAl alloy.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155706"},"PeriodicalIF":2.8,"publicationDate":"2025-02-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143535245","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":"Experimental study on the oxidation behavior of nuclear graphite IG-110","authors":"Bin Lin ,&nbsp;Liang Chen ,&nbsp;Guisen Liu ,&nbsp;Yao Shen","doi":"10.1016/j.jnucmat.2025.155712","DOIUrl":"10.1016/j.jnucmat.2025.155712","url":null,"abstract":"<div><div>This study presents an experimental investigation on the oxidation behavior of nuclear graphite IG-110. A gas chromatography-based oxidation test platform was developed to monitor the oxidation processes of IG-110 graphite in different atmospheres. The oxidation tests focused on the presence of H₂O and O₂, with temperatures up to 750 °C and various partial pressures. The graphite consumption was analyzed, and a Boltzmann function was proposed to describe the relationship between graphite consumption and oxygen partial pressure. A graphite consumption prediction model was developed based on this function, which can be utilized to calculate the graphite consumption of CO/CO₂ reaction products in High-Temperature Gas-cooled Reactors (HTGRs).</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155712"},"PeriodicalIF":2.8,"publicationDate":"2025-02-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143511579","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":"LBE erosion-corrosion behaviors of gelcasted high-speed rotating Ti3AlC2 impeller","authors":"Zhanchong Zhao ,&nbsp;Shibo Shi ,&nbsp;Xinxin Gao ,&nbsp;Qingsheng Wang ,&nbsp;Youyuan Zhang ,&nbsp;Yuying Wen ,&nbsp;Yang Ling ,&nbsp;Xian Zeng ,&nbsp;Mei Ma","doi":"10.1016/j.jnucmat.2025.155708","DOIUrl":"10.1016/j.jnucmat.2025.155708","url":null,"abstract":"<div><div>The MAX phase stands out as one of the highly prospective candidate materials for the impeller of the lead-bismuth fast reactor nuclear main pump. In order to determine the compatibility between the high-speed rotating Ti₃AlC₂ ceramic impeller and the liquid lead-bismuth eutectic (LBE) alloy, the corrosion behavior was comprehensively investigated within the LBE environment. In this study, the erosion-corrosion behavior of a high-speed rotating Ti₃AlC₂ ceramic impeller in LBE at 550 °C for 600 h was systematically investigated. The research results indicate that the Ti₃AlC₂ impeller maintained its structural integrity without macroscopic fractures after dynamic corrosion. Weight evaluation shows that during the corrosion test, the degradation of its physical or chemical properties was negligible. Surface analysis revealed the formation of a corrosion layer primarily composed of TiC and amorphous carbon, with PbO adhering to the surface as a protective barrier, effectively mitigating Ti and Al losses. Comparative analysis confirmed the superior adhesiveness of PbO over TiO₂ and Al₂O₃. The impeller maintained its mechanical performance, experiencing only a minor weight gain of 0.02 wt.% during the test. Vibration analysis confirmed operational stability, with a maximum stress of 10.76 MPa and a rotational frequency of 16.7 Hz, well below the first-order resonance frequency of 4997.7 Hz. This study furnishes crucial insights into the corrosion characteristics of the MAX phase and presents significant data. It is anticipated to provide a valuable reference for the initial application of MAX-phase ceramic impellers in advanced nuclear systems.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155708"},"PeriodicalIF":2.8,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143487305","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":"Microstructure, thermal properties and irradiation behaviors of uranium nitride (UN) nuclear fuel densified by Spark Plasma Sintering (SPS)","authors":"Jinlei Yang ,&nbsp;Jianjian Li ,&nbsp;Chao Yan ,&nbsp;Peng Wang ,&nbsp;Haobo Yang ,&nbsp;Xin Qian ,&nbsp;Kun Yang ,&nbsp;Shizhuan Xu ,&nbsp;Jun Lin","doi":"10.1016/j.jnucmat.2025.155709","DOIUrl":"10.1016/j.jnucmat.2025.155709","url":null,"abstract":"<div><div>This study investigated the microstructure, thermal properties, and irradiation behavior of uranium nitride (UN) fuel pellets synthesized via Spark Plasma Sintering (SPS). The highly densified UN pellets were successfully achieved by SPS at 1600 °C with a theoretical density above 96 % TD. The SPS-densified UN pellets exhibited good resistance to irradiation swelling, and the lattice constants, phase fractions, and surface roughness of the irradiated region demonstrated high stability. TEM results showed that dislocation loops and Ar bubbles were observed in the irradiated samples, with number densities of 4.74 × 10²³ and 1.06 × 10²⁴ /m³, respectively. Furthermore, the data analyzed by the Frequency Domain Thermal Reflection (FDTR) technique matched the data from Laser Flash Analyzer (LFA), confirming the suitability of FDTR for measuring the thermal conductivity of irradiated layers, which was found to be 20 % less than unirradiated values. It was inferred that the dislocation loops and Ar bubbles were the crucial reasons for the decreased thermal conductivity. The results provided valuable insights into the thermal properties and irradiation behaviors of SPS-sintered UN pellets.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"608 ","pages":"Article 155709"},"PeriodicalIF":2.8,"publicationDate":"2025-02-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143591960","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":"X-ray computed tomography of deconsolidated TRISO Particles from the AGR-5/6/7 irradiation experiment capsule 1 Compact","authors":"Rahul Reddy Kancharla ,&nbsp;William C. Chuirazzi ,&nbsp;Joshua J. Kane ,&nbsp;John D. Stempien ,&nbsp;Cameron Howard ,&nbsp;Swapnil Morankar ,&nbsp;Miles T. Cook ,&nbsp;Quintin D. Harris","doi":"10.1016/j.jnucmat.2025.155704","DOIUrl":"10.1016/j.jnucmat.2025.155704","url":null,"abstract":"<div><div>X-ray Computed Tomography (XCT) was performed on TRISO particles from fuel Compact 1-7-9 from Capsule 1 of the AGR-5/6/7 irradiation experiment. Compact 1–7–9 was in the vicinity of overheated thermocouples that released transition metals and caused significant damage and failure of TRISO particles within regions of Capsule 1. Four particles, Particles A–D, were non-destructively examined in three-dimensions (3D). Particle B was found to have degraded silicon carbide (SiC) and significant amounts of Ni, actinides, and fission products that contributed to this degradation. The nickel (Ni) showed some evidence of penetration into Particle B from the exterior surface of the silicon carbide layer. The kernels of Particles A, C, and D were analyzed for their size, shape, and observed porosity. A density gradient was observed in the kernels, which may be nascent kernel migration observed in fuel with higher irradiation temperatures. In addition to the kernels, Particles A, C, and D possessed large clusters of dense material that were grouped primarily to one side of the particle located at the buffer/inner pyrolytic carbon interface. These clusters were attributed to nickel migration into the particle interiors, which was confirmed with scanning electron microscopy (SEM) and electron dispersive X-ray spectroscopy (EDS).</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"607 ","pages":"Article 155704"},"PeriodicalIF":2.8,"publicationDate":"2025-02-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143478884","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}