Nuclear Engineering and Design最新文献

{"title":"Evaluation of fully developed turbulent friction coefficient in polygonal section ducts and ducts with corner angle 0° by new characteristic lengths","authors":"Son Nam-Jin,&nbsp;Kim RyongIl,&nbsp;Kim Hyo-Song,&nbsp;Yun Kumchol,&nbsp;O. Ju-Sung","doi":"10.1016/j.nucengdes.2024.113663","DOIUrl":"10.1016/j.nucengdes.2024.113663","url":null,"abstract":"<div><div>In estimating the friction coefficient of fully developed turbulent flow in a circular duct, the Blasius equation is often used. The hydraulic diameter is the characteristic length of the <em>Re</em> number used when mapping the non-circular duct section to the circular pipe section. We have newly defined the characteristic length of the <em>Re</em> number of the Blasius equation. The newly proposed characteristic length is wetted perimeter equivalent round diameter and the modified hydraulic diameter with minimum change of area and perimeter. Based on the new diameters, a method to estimate the turbulent friction coefficient of a fully developed isosceles.</div><div>triangular duct, regular polygon duct, and a duct with an edge angle of 0° is proposed and its accuracy is evaluated. The new characteristic lengths we propose are easy to calculate, provided that certain accuracy is achieved. Therefore, it is easily applicable to the estimation of friction coefficient of isosceles triangular ducts, regular polygonal ducts and corner angle 0° section ducts.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142593935","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":"Closing the nuclear fuel cycle: Strategic approaches for NuScale-like reactor","authors":"Keferson de A. Carvalho ,&nbsp;Graiciany Barros ,&nbsp;Matheus H.S. Araújo ,&nbsp;Andre A. Campagnole dos Santos ,&nbsp;Vitor Silva ,&nbsp;Tiago Augusto Santiago Vieira ,&nbsp;Rebeca Cabral Gonçalves","doi":"10.1016/j.nucengdes.2024.113672","DOIUrl":"10.1016/j.nucengdes.2024.113672","url":null,"abstract":"<div><div>The present study proposes the potential implementation of eight different closed fuel cycle strategies for a NuScale-like reactor core using its own spent fuel as a reusable source of fissile material for energy production. For that, the spent fuel composition after three burnup cycles of approximately 12 MWd/kgU of NuScale-like initial core and five years of cooling in a spent fuel pool was theoretically reprocessed by GANEX or UREX+ methods. After reprocessing, these two new fuel compositions were spiked in a mixture of thorium (Th) or depleted uranium (DpU), and afterwards inserted into specific batch positions of the core. Therefore, the proposed NuScale-like core configurations contain fuel assemblies loaded with conventional uranium-based fuel and others loaded with reprocessed fuel, resulting in the following combinations: UO₂ and GANEX spiked with Th, UO₂ and GANEX spiked with DpU, UO₂ and UREX+ spiked with Th, UO₂ and UREX+ spiked with DpU. The main idea is to comprehend the advantages of adopting the closed nuclear fuel cycle for a NuScale-like reactor by comparing the reference case and the cases containing reprocessed fuel. The results exhibited that all instances in which the core was simulated with reprocessed fuel improved the feedback coefficient, maximum excess of reactivity varying the boron concentration in the coolant, and power peak factor (PPF). Furthermore, the closed nuclear fuel strategies also demonstrated savings of about 17.50% in terms of separating work units (SWU) due to plutonium and uranium recycling, and a potential burnup extension of approximately 43%. The Serpent code version 2.1.32 developed by VTT and ENDF/B-VII.0 nuclear data library has been used to perform the simulations.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142593939","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":"Seismic behavior analysis of control rod dropping by vector form Intrinsic Finite-Element method","authors":"Han Wu ,&nbsp;Qing Peng ,&nbsp;Fenglei Jin ,&nbsp;Jingru Song ,&nbsp;Xiaoming Liu","doi":"10.1016/j.nucengdes.2024.113660","DOIUrl":"10.1016/j.nucengdes.2024.113660","url":null,"abstract":"<div><div>Control rod drive mechanism (CDRM) play a major role in ensuring safe operation of nuclear reactor during the earthquake, under which the dropping time of control rod is crucial for safe shutdown. Under the earthquake, Rod Cluster Control Assembly (RCCA) has contact collision with the guide tube, resulting in an increase of friction and a decrease of the speed of the falling rod. In addition, in view of the slender structure of the falling rod, the flexible deformation vibration will occur under the impact excitation, which will aggravate the collision and friction. In order to solve the nonlinear problem caused by contact collision between control rod and guide tube, we proposed a dynamic behavior analysis program of rod dropping based on vector finite element method. In order to simulate contact collision force more accurately, we proposed a conformal contact law to simulate the contact force between control rod and guide tube. The vector finite element model and simulation program are validated by comparing with a rod drop experiment. Based on the developed program, the rod dropping behavior, including rod dropping time, contact force, friction force and control rod deformation under several earthquake condition were discussed in this work.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142593938","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":"Stability and bifurcation analysis of Oskarshamn-2 event with nuclear data and kinetic parameter uncertainties","authors":"A. Dokhane,&nbsp;A. Vasiliev,&nbsp;H. Ferroukhi","doi":"10.1016/j.nucengdes.2024.113678","DOIUrl":"10.1016/j.nucengdes.2024.113678","url":null,"abstract":"<div><div>The primary aim of the current research is to investigate the impact of nuclear data and kinetic parameters uncertainties on the stability and bifurcation behaviour of the Oskarshamn-2 core under unstable conditions. Utilizing the SHARK-X platform, nuclear data and kinetic parameter uncertainties are propagated in 2-D lattice calculations with CASMO5 and downstream to static and dynamic calculations using SIMULATE3 and SIMULATE-3 K. Results show that, the nuclear data uncertainties have a drastic effect on the stability behaviour of the core when the instability is triggered because the system become highly nonlinear at such conditions. However, more interesting is the qualitative effect on the stability behaviour where the nature of solution changes, i.e. occurrence of bifurcation, from a highly unstable state (diverging oscillation amplitudes) to a highly stable state (rapidly decreasing oscillation amplitudes). This change in the nature of behaviour, i.e. solution, is found to be due to the fact that the stability event occurs very close to the stability boundary of the system and therefore any change in any parameter could be enough to swing the system to the other side of the stability boundary. Concerning kinetic parameters, results show a clearly smaller impact compared to that of nuclear data, leading to uncertainties in the decay ratio and resonance frequency around 2.5 % and 0.2 % respectively. The main effect is variations in oscillation amplitude without altering the nature of the solution.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-11-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142586624","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":"MELCOR 2.2 iPWR LOCA type accident analysis, PART II: BDBA","authors":"M. Malicki,&nbsp;T. Lind","doi":"10.1016/j.nucengdes.2024.113667","DOIUrl":"10.1016/j.nucengdes.2024.113667","url":null,"abstract":"<div><div>The integrated Pressurized Water Reactor brings many potential improvements for the nuclear industry, such as passive and modular design, which potentially supports reliability and safety. Besides their advantages, passive systems are also more challenging to simulate, and predictions of the complex behavior of the reactor, especially under accident conditions, need to be validated. In the first part of this study, the authors analyzed the thermal-hydraulic response of the iPWR to several design basis accident sequences without entering into the severe accident domain. As a continuation of the investigation, a sensitivity study of the beyond design basis accident scenario was performed and analyzed as described and presented here, part 2 of the study.</div><div>A generic iPWR MELCOR 2.2 input deck was developed and used to perform a loss-of-coolant accident (LOCA)-type scenario analysis in which a break is assumed in the chemical and volume control system line. The effect of the elevation of the break and decay heat on accident progression is investigated. This allows the examination of input deck and code reliability under different conditions, from full core uncovery to mitigated accidents. Overall, eight cases were calculated in which the break elevation and decay heat were varied, providing knowledge about the modeling of the iPWR design and potential analytical challenges, which was the main goal of this work.</div><div>The analyses show that MELCOR 2.2 can model iPWR design and simulate severe accident scenarios with different levels of core degradation. One of the technical <u>insights</u> from this preliminary study was that natural circulation plays a significant role in the late phase of a severe accident when the core is uncovered.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-11-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142579013","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":"Impact of anisotropy on TRISO fuel performance","authors":"Gyanender Singh ,&nbsp;Jordan A. Evans ,&nbsp;Wen Jiang ,&nbsp;Jason Hales ,&nbsp;Stephen Novascone","doi":"10.1016/j.nucengdes.2024.113637","DOIUrl":"10.1016/j.nucengdes.2024.113637","url":null,"abstract":"<div><div>Manufacturing of tristructural isotropic (TRISO) particles involves the deposition of pyrolytic carbon (PyC) and silicon carbide (SiC) layers using the fluidized bed chemical vapor deposition (CVD) process. The CVD process is known to generate polycrystalline layers with crystallographic textures, which imparts anisotropic thermophysical properties to the layers. Past studies have shown the risk for particle failure increases with an increase in anisotropy. The limit beyond which the anisotropy of PyC layers becomes unacceptable due to failure risk has been identified as a high-priority knowledge gap. This work presents a first systematic study on the effects of anisotropic thermal and mechanical properties on TRISO fuel performance. This computational study, performed using the fuel performance code BISON, investigates how the anisotropy in elasticity and thermal properties affect the stresses, temperature, and failure of a TRISO particle. The influence of other factors, such as operating temperature and particle geometry on the anisotropy effects, also has been analyzed. The studies utilize the recently published anisotropic elasticity and thermal behavior models for TRISO PyC and SiC layers implemented using tensors with full anisotropic capability. The spherical TRISO particles with anisotropic properties were found to have greater maximum tensile stress and significantly higher failure probability than the spherical particles with isotropic properties. The fuel performance predicted using these recently developed models was found to be comparable with the performance obtained using the historical models.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-11-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571567","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":"International fuel performance study of fresh fuel experiments for PCMI effects during RIA experiments","authors":"Seokbin Seo ,&nbsp;Charles Folsom ,&nbsp;Colby Jensen ,&nbsp;David Kamerman ,&nbsp;Luana Giaccardi ,&nbsp;Marco Cherubini ,&nbsp;Pavel Suk ,&nbsp;Martin Sevecek ,&nbsp;Jerome Sercombe ,&nbsp;Isabelle Guenot-Delahaie ,&nbsp;Alessandro Scolaro ,&nbsp;Matthieu Reymond ,&nbsp;Katalin Kulacsy ,&nbsp;Luis Herranz ,&nbsp;Francisco Feria ,&nbsp;Pau Aragón ,&nbsp;Grigori Khvostov ,&nbsp;Imran Khan ,&nbsp;Anuj Kumar Deo ,&nbsp;Srinivasa Rao Ravva ,&nbsp;Carlo Fiorina","doi":"10.1016/j.nucengdes.2024.113673","DOIUrl":"10.1016/j.nucengdes.2024.113673","url":null,"abstract":"<div><div>This paper presents the results of High-burnup Experiments for Reactivity-initiated Accident (HERA) Modeling &amp; Simulation (M&amp;S) exercise. The HERA project under the Nuclear Energy Agency (NEA) Second Framework for Irradiation Experiments (FIDES-II) program is focused on studying Light Water Reactor (LWR) fuel behavior during Reactivity-Initiated Accident (RIA) conditions. The Part I M&amp;S cases are based on a series of tests in the Transient Reactor Test (TREAT) facility in the United States and the Nuclear Safety Research Reactor (NSRR) in Japan. The purpose of this work is to evaluate the test design to accomplish its goals in establishing clearer understanding of the effects of power pulse width during RIA conditions. The blind predictions using various computational tools have been performed and compared amongst to interpret the behaviors of high burnup fuels during RIA. While many international participants evaluate the thermal–mechanical behavior of fuel rod under different conditions, a considerable scatter of outputs comes out for the cases due to the disparity between codes in predicting mechanical behaviors. In general, however, the results of thermal–mechanical analysis elaborate that nominal design conditions the shorter pulse width tests in NSRR should cause cladding failures while the TREAT tests appear to have more split prediction of failure or not. Furthermore, the sensitivity analysis varying key testing parameters reveals the considerable effect of power pulse width and total energy deposition on prediction of fuel rod failure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-11-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571639","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":"Rapid deep learning prediction model using satellite imagery for radiation accident Announcement system in Serbia","authors":"Slavko Dimović,&nbsp;Milica Ćurčić,&nbsp;Dušan Nikezić,&nbsp;Ivan Lazović,&nbsp;Dušan Radivojević","doi":"10.1016/j.nucengdes.2024.113657","DOIUrl":"10.1016/j.nucengdes.2024.113657","url":null,"abstract":"<div><div>Radioactivity environmental monitoring with the help of the Radiation Accident Announcement System (RAAS) established in the Republic of Serbia is of vital importance for rapid response in the event of intervention dose values being reached such as Operational Intervention levels (OILs). There are cases of impossibility of using a ground-based model in order to predict the transport and spread of radiation during a nuclear accident such as the one in the Fukushima Daiichi nuclear power plant 2011. Modern technology has made it possible to use machine learning models with remote sensing data in addition (or an alternative) to atmospheric models with ground data collection methods. Deep learning (DL) model was developed and trained on a satellite-based cloud fraction dataset to forecast diurnal cloud drift. By forecasting the movement of clouds that are potential carriers of radioactive materials, decision-makers can provide an adequate response with an Action Plan in the case of nuclear and radiation accidents and incidents. Designed Application Programming Interface (API) has been developed and integrated between the DL model and the RAAS. In this way, the monitoring, data sharing and exchange processes were automated in order to trigger the DL model when an OILs value of 1 μSv/h is reached. The hyperparameter optimization process of the DL model was done with grid search and Particle Swarm Optimization (PSO) to achieve maximum performance on the data in a reasonable amount of time. The evolution metrics to monitor and measure the performance of the DL model were cosine similarity (CS) and structural similarity (SS) with the best score of 0.78282 and 0.27063 for grid search, respectively, while 0.27065 and 0.78282 for PSO, respectively. It can be concluded that remote sensing imagery with DL model is an alternative approach against a ground-based forecasting system, and is able to predict in near real-time independently of ground network system and data.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-11-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142571565","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":"Feasibility analysis of Uranium-Based fission fragment magnetic collimation space propulsion system","authors":"Dacai Zhang,&nbsp;Longyu Fan,&nbsp;Xirui Zhang,&nbsp;Guanghui Zhong,&nbsp;Ganglin Yu,&nbsp;Kan Wang","doi":"10.1016/j.nucengdes.2024.113674","DOIUrl":"10.1016/j.nucengdes.2024.113674","url":null,"abstract":"<div><div>The theoretical specific impulse of the direct fission fragment propulsion system exceeds 10⁶ s, making it more appropriate for long space exploration compared to existing nuclear propulsion methods. Previous studies on fission fragment propulsion systems have been limited to employing Am-242 m as fuel, lacking investigations on system reactivity control and fuel temperature distribution. This study suggested a direct fission fragment propulsion system using U-235 as fuel. Firstly, the neutronics features of the system were investigated using the Monte Carlo software RMC. Subsequently, Fluent was used to compute the temperature distribution of the fuel assembly. MATLAB was then applied to analyze the temporal variation of the system’s propulsion properties. Finally, “navigation cost” was defined to compare the performance of different propulsion methods. Computation results revealed that the uranium-based fission fragment propulsion system exhibits good critical and safety features. The initial k_eff of the system is 1.015, the specific impulse reaches 5.88 × 10⁴ s, and the fuel element temperature is below 1000 K. Additionally, the navigation cost of the fission fragment propulsion system is lowered by 2 to 6 orders of magnitude compared to other propulsion technologies. This study validates the theoretical viability of the uranium-based fission fragment propulsion system and demonstrates its significant advantages for long-distance deep space exploration missions, providing an alternate option for deep space exploration.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142560905","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":"Toward development of a low-temperature failure envelope of cases for high-burnup RIAs under PWR operational conditions","authors":"L. Aldeia Machado ,&nbsp;K. Nantes ,&nbsp;E. Merzari ,&nbsp;L. Charlot ,&nbsp;A. Motta ,&nbsp;W. Walters","doi":"10.1016/j.nucengdes.2024.113642","DOIUrl":"10.1016/j.nucengdes.2024.113642","url":null,"abstract":"<div><div>A Reactivity-Initiated Accident (RIA) is a design-basis accident that occurs when the reactor loses one of its control rods. A reactivity insertion will follow such events, drastically increasing the fuel pellet’s temperature and volume due to thermal expansion. The fuel pellet and the cladding will interact mechanically, which could lead to cladding failure. This work presents the development of cases where low-temperature failures are more likely to happen for high-burnup fuels under PWR operational conditions. A coupled computational model between the nuclear fuel performance code BISON, MOOSE’s Thermal-Hydraulic Module (MOOSE-THM), and MOOSE’s Stochastic Tools Modules (MOOSE-STM) was created to study the thermal-hydraulic behavior of a high-burnup fuel rodlet during the first stages of an RIA transient, allowing us to identify three scenarios: the system reached CHF, leading to high-temperature failure, the system failed due to PCMI, or the system survived the whole transient without failing. To address these scenarios, we performed a sensitivity analysis with more than 140,000 model replicates through MOOSE-STM varying parameters such as the power pulse width, power pulse total energy deposition, hydride rim thickness, coolant inlet temperature, and coolant inlet mass flux. We also compared two PCMI failure criteria in our analysis. Our results suggest that the hydride rim thickness and the power pulse width will be the key parameters impacting the failure type our system would undergo during the power transient. Using the data from our simulations, we constructed two failure maps, one for each PCMI failure criterion, showing how the failure type is affected by each parameter considered in the sensitivity analysis. We also provided a closed-form expression for the boundary between the PCMI and CHF failure types as a function of the hydride rim thickness and power pulse width.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142552531","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}