Progress in Nuclear Energy最新文献

{"title":"Study on flow heat transfer characteristics and mechanism of lithium-lead and helium-xenon","authors":"Zhuoran Lei ,&nbsp;Ruibo Lu ,&nbsp;Xiaojian Ge ,&nbsp;Haoran Shen ,&nbsp;Fulong Zhao ,&nbsp;Sichao Tan ,&nbsp;Ruifeng Tian","doi":"10.1016/j.pnucene.2025.105774","DOIUrl":"10.1016/j.pnucene.2025.105774","url":null,"abstract":"<div><div>To find out the coupling heat transfer problem between liquid lithium-lead (Li-Pb) and helium-xenon (He-Xe) in a Printed Circuit Heat Exchanger (PCHE), a three-dimensional model of a double-straight-through PCHE heat exchange unit was established. This paper employs numerical simulation to investigate the coupled heat transfer characteristics as well as the mechanism between liquid lithium-lead and helium-xenon and to fit the flow heat transfer correlation of lithium-lead under specific working conditions. The results show that the coupled heat transfer of liquid lithium-lead is mainly dominated by thermal conduction, mainly influenced by the wall temperature in contact with the fluid and the heat flux density at the wall. Convective heat transfer is the main mechanism influencing helium-xenon's coupled heat transfer. Modifying the inlet velocity of helium-xenon has a more significant effect on the heat exchanger's performance compared to adjusting the inlet velocity of liquid lithium-lead. Regardless of which side's temperature is changed, the heat transfer amount will be greatly affected, and the heat exchanger's overall heat transfer coefficient is mainly determined by the helium-xenon.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105774"},"PeriodicalIF":3.3,"publicationDate":"2025-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143760302","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":"Non-condensable gas accumulation and distribution due to condensation in the CIGMA facility: Implications for Fukushima Daiichi Unit 3 (1F3)","authors":"Ari Hamdani ,&nbsp;Shu Soma ,&nbsp;Satoshi Abe ,&nbsp;Yasuteru Sibamoto","doi":"10.1016/j.pnucene.2025.105771","DOIUrl":"10.1016/j.pnucene.2025.105771","url":null,"abstract":"<div><div>This study was motivated by previous analysis by TEPSYS, which indicated that different temperatures in the 5th and 4th floors of the reactor building (R/B) of Fukushima Daiichi Unit 3 (1F3) could significantly influence the distribution of non-condensable gases during the severe accident in 2011. Understanding this effect is crucial for assessing the risks associated with non-condensable gas accumulation in R/B, particularly considering that the hydrogen explosion in 1F3 may not have been a single-stage event but could have involved a multi-stage explosion. To investigate the distribution and transport on non-condensable gas under varying cooling conditions, an experimental study was conducted using the CIGMA facility, a large-scale test vessel with segmented sections created by partition plates, designed to replicate the structure and conditions of the R/B. Steam and helium, used as non-condensable hydrogen substituted, were continuously injected at the top of the CIGMA vessel for 10,000 s to replicate the leakage of steam and hydrogen through the shield plug. Two cooling conditions were tested: Case 1 with a cooling temperature of 50 °C, and Case 2 with a cooling temperature of 90 °C. The experiments aimed to explore how varying cooling temperatures impact helium accumulation in different regions of the R/B structure. The study found that in both cases, the highest concentration of non-condensable gases may not always be near the injection point but rather in downstream regions. In these regions, steam and non-condensable gases move downward and continue to condense, significantly enhancing the concentration of helium. Specifically, in Case 1, after 10,000 s, the helium concentration reached 65 % in the middle region (representing the 4th floor of R/B) and 45 % in the top region (representing the 5th floor of R/B). Analysis using the Shapiro ternary diagram showed that, under these conditions, the gas mixture in the middle region posed a potential detonation risk. This study provides important insights into gas distribution within a nuclear reactor building during severe accident conditions, which are crucial for developing more effective safety measures and risk mitigation strategies in nuclear reactor designs.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105771"},"PeriodicalIF":3.3,"publicationDate":"2025-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143760301","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":"Evaluation of load following control strategy for heat pipe-cooled microreactor coupled with closed Brayton cycle system","authors":"Yiran Qian,&nbsp;Tao Liu,&nbsp;Yan Zhang,&nbsp;Yuyang Leng,&nbsp;Tianrun Shi,&nbsp;Meihui Song,&nbsp;Weixiong Chen","doi":"10.1016/j.pnucene.2025.105768","DOIUrl":"10.1016/j.pnucene.2025.105768","url":null,"abstract":"<div><div>Heat pipe-cooled reactors (HPCRs), with their straightforward and compact design and inherent safety features, are considered a promising option for mobile nuclear power sources. However, the coupling between the HPCR and the energy conversion system complicates load tracking control. In this study, a comprehensive dynamic model of an HPCR coupled with a closed-air Brayton cycle (CABC) is developed. This model integrates a neutron kinetics point reactor model, a thermal resistance network model for the heat pipe, and a CABC model representing the power conversion system. To achieve effective load tracking, a control scheme for the Brayton cycle power is proposed, combining turbine bypass control with inventory control. The system's response to load variations, from 100 % full power (FP) to 50 % FP at a rate of 5 % FP/min, is analyzed, considering reactor power self-regulation. The results indicate that the reactor power exhibits an overshoot of −4.36 % and stabilizes after 2100s, while the fuel average temperature fluctuates within ±4.6 °C and stabilizes at 2240s. Building on the dynamic performance of the HPCR-CABC system, a reactor power control scheme is developed by adjusting control drums to introduce external reactivity. This scheme successfully controls reactor power at a steady state within 720s, without fluctuation. However, the fuel average temperature fluctuates by more than 10 °C, stabilizing after 9450s. To improve control effectiveness, a temperature correction channel is incorporated into the reactor power control system. As a result, the fuel average temperature fluctuates within ±0.3 °C and stabilizes at the design value of 699.4 °C within 860s. Compared to previous control schemes, the improved cascade reactor power control scheme results in a shorter stabilization time for the fuel average temperature. To assess the adaptability of the proposed HPCR-CABC coordination control strategy, simulations are conducted for load reductions from 100 % FP to 80 %, 60 %, and 40 % FP at a variation rate of 20 % FP/min. The results demonstrate that the proposed HPCR-CABC coordination control strategy exhibits strong load-following capability.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105768"},"PeriodicalIF":3.3,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143738250","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":"Investigation and optimization on performance of an innovative Brayton cycle in space nuclear power system","authors":"Tong Lu ,&nbsp;Ersheng You ,&nbsp;Haochun Zhang ,&nbsp;Fangwei Ma","doi":"10.1016/j.pnucene.2025.105766","DOIUrl":"10.1016/j.pnucene.2025.105766","url":null,"abstract":"<div><div>As space exploration advances, the need for robust and dependable energy sources continues to grow. Due to its exceptional energy density, substantial output power, prolonged operational lifespan, and remarkable adaptability to environmental conditions, the space nuclear power system (SNPS) stands out as the ideal energy solution for future deep-space missions. At the same time, the Brayton cycle is an ideal thermoelectric conversion method for megawatt SNPS. However, for the Brayton system, the circulating working fluid still has high energy quality after heating by the reactor and doing work by the turbine. And the compressor is the main power consumption equipment in the Brayton system, it is necessary to reduce its inlet temperature as much as possible. In this paper, under the background of SNPS, a multi-compression Brayton cycle was proposed. Based on the simple Brayton cycle structure, the circulating working fluid was bled and multiple compressed. Utilizing He-Xe mixture as the working fluid, thermodynamic and mass estimation models were developed to analyze how bleeding ratio, turbine inlet temperature, recuperator effectiveness, compressor inlet temperature and pressure ratio, influence the thermal efficiency, overall mass, and specific power of the multi-compression Brayton system. Taking thermal efficiency, total mass and specific power of the system as optimization objectives, the non-dominated sorting whale optimization algorithm (NSWOA) was used for multi-objective optimization. Under the compromise of maximum thermal efficiency, minimum total mass and maximum specific power, the optimal value of system thermal efficiency, total mass, specific power was 61.04 %, 1266.53 kg and 0.31 kWe/kg, respectively.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105766"},"PeriodicalIF":3.3,"publicationDate":"2025-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143738249","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":"Shielding optimization of a heat pipe cooled reactor based on shielding material evaluation","authors":"J.C. Huang,&nbsp;B.H. Yan,&nbsp;L.S. Wang","doi":"10.1016/j.pnucene.2025.105763","DOIUrl":"10.1016/j.pnucene.2025.105763","url":null,"abstract":"<div><div>The miniaturization of the shielding layer plays a crucial role in the reactor's design. To ensure the reactor is more portable and adaptable to flexible scenarios, the shielding layer of the reactor is optimized. Different shielding materials are evaluated by using a series of parameters in this work, and a material evaluation form is obtained as a result. This evaluation form is proved to have guiding significance for material selection by simulation of the reactor. Based on the evaluation form, optimized shielding materials are selected for further optimization. By using the selected materials of the evaluation form, the mass and volume of the shielding layer have been reduced. Finally, the shielding layer is redesigned based on the previous optimization, which shows a better shielding effect. The weight reduction can reach 22 % compared to the previous design at the end of optimization. This means the heat pipe reactor can be miniaturized to meet the needs of more complex situations.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105763"},"PeriodicalIF":3.3,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143725987","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":"Analysis of electroneutrality of fission products in molten salt reactors","authors":"Seokjoo Yoon,&nbsp;Jun Woo Park,&nbsp;Hyeonjin Eun,&nbsp;Eunhyug Lee,&nbsp;Yonghee Kim,&nbsp;Jong-Il Yun","doi":"10.1016/j.pnucene.2025.105764","DOIUrl":"10.1016/j.pnucene.2025.105764","url":null,"abstract":"<div><div>Liquid-fuel molten salt reactors are a type of Generation IV reactors that has attracted attention for their enhanced safety and thermal efficiency. However, the liquid state of the fuel introduces complexity to the reactor system, particularly in relation to the behavior of fission products and the challenge of maintaining electroneutrality. The interaction of the liquid fuel and the container material, along with the release of noble gas fission products, complicates the analysis of electroneutrality. This study examines the electroneutrality of fission products from key fissile isotopes that could be utilized in molten salt reactors. The calculations take into account variables such as the type of salt (chlorides or fluorides), fissile isotope (Th, U, or Pu), and neutron energy (thermal or fast). Three scenarios involving noble gas release are compared to assess its impact on electroneutrality. Fission of Th-232, U-233, and U-235 can mitigate oxidizing conditions post-fission, by controlling the initial valence state of the fissile isotopes. In contrast, the fission of U-238 and Pu isotopes consistently results in an oxidizing environment. These findings indicate that molten salt reactors relying solely on U-238 and Pu may experience oxidizing conditions due to the fission reaction, suggesting that additional measures may be necessary to prevent excessive oxidation of the reactor system.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105764"},"PeriodicalIF":3.3,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143735086","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 comparative study and design optimization of potential cladding materials for gas cooled fast reactor: Neutronics, radiation shielding, and thermal hydraulics analysis","authors":"Shohanul Islam ,&nbsp;Jehad Mohammed Haque","doi":"10.1016/j.pnucene.2025.105741","DOIUrl":"10.1016/j.pnucene.2025.105741","url":null,"abstract":"<div><div>This study presents a comparative analysis of potential cladding materials for gas-cooled fast reactors (GFR), evaluating neutronics, radiation shielding, and thermal hydraulics, and includes optimized mixed cladding design to improve neutronics and overall performance. Eight different cladding materials—SiC, HT9, ODS steel, 15-15Ti, SS 316, TZM alloy, Ta alloy, and Zircaloy-4 were studied. SiC cladding achieved the longest cycle length, with Zircaloy-4 following closely, while metallic models generally showed shorter cycles, except for the optimized TZM mixed cladding, which also achieved extended cycle lengths. Though the optimized Ta alloy model did not achieve a longer cycle length, it improved radial neutron flux distribution and reduced radial power peaking factors, with TZM model contributing similarly. The study also found that SiC, TZM, and Ta alloy models produced harder neutron spectra and demonstrated reduced fissile inventory depletion and enhanced minor actinide transmutation. SiC and TZM model had the smallest reactivity swings, with all models except HT9 and ODS steel showing satisfactory beta effective values, and all models except Zircaloy-4 exhibited acceptable negative Doppler constant values. Radiation shielding assessments revealed SiC had the highest HVL, TVL, and MFP, while TZM and Ta alloy had the lowest, indicating better shielding performance. Thermal hydraulics analysis showed that these two materials exhibited lower peak temperatures in both fuel and cladding, along with a more uniform temperature distribution compared to other models. Despite these differences, all models maintained an adequate thermal margin in the cladding region.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105741"},"PeriodicalIF":3.3,"publicationDate":"2025-03-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143714956","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":"Induction heating of composite billet of T91 and Zircaloy-4 for manufacturing of double clad tube: Simulation and experimental validation","authors":"Navinkumar Dubey ,&nbsp;Aarti Chaudhary ,&nbsp;Arijit Laik ,&nbsp;Sugilal Gopalakrishnan ,&nbsp;Komal Kapoor","doi":"10.1016/j.pnucene.2025.105761","DOIUrl":"10.1016/j.pnucene.2025.105761","url":null,"abstract":"<div><div>Metallic fuel is the preferred choice for future Fast Breeder Reactors due to higher breeding ratio. A mechanical bonded double clad tube with a T91 steel outer shell and Zr inner liner is a potential choice for metallic fuel. However, the metallurgical bonded clad tube will have superior properties against the mechanical bonded tube. The metallurgically bonded co-axial tube of T91 and Zr/Zircaloy-4 (Zr-4) can be manufactured by the co-extrusion route. The significant difference in deformation behavior of the two alloys demands a temperature gradient over the thickness of the composite billet for co-extrusion. The fine control of temperature gradient across the cross section of the billet can be achieved only by induction heating. The present study focuses on simulating the induction preheating process of a double-clad billet of outer T91 and inner Zr-4. A 2-D axisymmetric model was considered and simulation studies with several key parameters such as coil frequency, current, pitch, and diameter were carried out with an aim to reduce temperature variation at the interface of T91 and Zr-4 in the billet. Simulations using optimised coil frequency and current reduced the temperature variation at the interface to as low as 22 °C. Subsequent simulations with variable coil pitch and diameter reduced the temperature variation to less than ±3 °C at the interface ignoring axial ends. The mean temperature across the T91 was 967.4 °C and across Zr-4 was 850.3 °C. Experiments were conducted for the co-axial tube of T91 and Zr-4 to validate the simulation model.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105761"},"PeriodicalIF":3.3,"publicationDate":"2025-03-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143725986","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":"Enhancing nuclear power plant diagnostics: A comparative analysis of XAI-based feature selection methods for abnormal and emergency scenario detection","authors":"Young Ho Chae ,&nbsp;Seung Geun Kim ,&nbsp;Jeonghun Choi ,&nbsp;Seo Ryong Koo ,&nbsp;Jonghyun Kim","doi":"10.1016/j.pnucene.2025.105759","DOIUrl":"10.1016/j.pnucene.2025.105759","url":null,"abstract":"<div><div>This study introduces the application of explainable artificial intelligence (XAI) techniques to enhance nuclear power plant diagnostics through effective feature selection. We compared various XAI methods, including gradient-based techniques, layer-wise relevance propagation, DeepSHAP, integrated gradients, local interpretable model-agnostic explanation(LIME), and saliency maps, with traditional approaches such as principal component analysis (PCA). By applying these methods to data from an IAEA iPWR simulator, which includes 35 abnormal and emergency scenarios with 116 state variables, we demonstrated the superiority of XAI-based methods in selecting features that effectively distinguish between different plant conditions. Our approach successfully reduced the input dimensionality from 116 to 20 features while maintaining high diagnostic accuracy. XAI methods, particularly saliency map and DeepSHAP, outperformed traditional techniques by revealing distinct patterns for various abnormal situations. This reduction in dimensionality offers several benefits, including enhanced cybersecurity, improved human–machine interfaces, and increased computational efficiency. The findings have significant implications for developing more accurate, efficient, and interpretable diagnostic systems in nuclear power plants, potentially improving safety and operational effectiveness. Future work will focus on validating these methods across diverse plant designs and integrating this approach with advanced AI techniques for real-time adaptive diagnostics.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105759"},"PeriodicalIF":3.3,"publicationDate":"2025-03-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143704338","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"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":"Influence of healing time, technological void and temperature on the shear behavior of the GMZ bentonite healing interface","authors":"Yi-ran Zhang ,&nbsp;Yong-gui Chen ,&nbsp;Yu-cheng Li ,&nbsp;Wei-min Ye ,&nbsp;Qiong Wang ,&nbsp;Zhao Sun","doi":"10.1016/j.pnucene.2025.105757","DOIUrl":"10.1016/j.pnucene.2025.105757","url":null,"abstract":"<div><div>Bentonite is an engineering barrier in the deep geological repository (DGR), and its mechanical properties are negatively impacted by the presence of technological voids. Based on direct shear test and mercury intrusion porosimetry (MIP) technique, the effects of healing time, technological void and temperature on the healing interface shear behavior of Gaomiaozi (GMZ) bentonite were studied. The results indicate that the shear strength of healing interface is directly related to the dry density of the interface, which is influenced by healing time, technological void and temperature. Within the healing time range studied, as healing time increased, bentonite continued to swell, leading to a rise in the dry density of the interface and an enhancement in shear strength. During this process, constrained by the limited swelling space, the intra-aggregate porosity decreases at first and then increases until stabilizing, while the inter-aggregate porosity is on the contrary, which reduces the distance between particles, thereby increasing the cohesion and friction angle of the interface. When the initial void of the interface increases, the shear strength at the interface significantly decreases, with corresponding reductions in cohesion and friction angle. As the temperature increases, the shear strength increases, notably in cohesion, while the friction angle shows no clear pattern of change. This is primarily attributed to the rearrangement of bentonite particles, the reformation of interlayer physical and chemical bonds, and the reduced viscosity of pore fluids. The findings of this study provide important insights into the mechanical stability of engineered barriers in the DGR.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105757"},"PeriodicalIF":3.3,"publicationDate":"2025-03-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143715068","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}