Nuclear Engineering and Design最新文献

{"title":"Integration of machine learning models for enhancing radioactive waste management of disused sealed radioactive sources","authors":"Ihsan Aulia Rahman ,&nbsp;Zico Pratama Putra ,&nbsp;Pendi Rusadi ,&nbsp;Kanita Salsabila Dwi Irmanti ,&nbsp;Ajrieh Setyawan ,&nbsp;Moch Romli ,&nbsp;Ayi Muziyawati ,&nbsp;Suhartono Suhartono ,&nbsp;Hendra Adhi Pratama ,&nbsp;Raden Sumarbagiono ,&nbsp;Gustri Nurliati ,&nbsp;Niken Siwi Pamungkas ,&nbsp;Muhammad Yusuf","doi":"10.1016/j.nucengdes.2025.114272","DOIUrl":"10.1016/j.nucengdes.2025.114272","url":null,"abstract":"<div><div>This research presents a novel machine-learning approach for optimizing the radioactive waste management of Disused Sealed Radioactive Sources (DSRS) through advanced predictive modelling. The study leverages comprehensive data from the Radioactive Waste Treatment Facility (IPLR), combining 1,339 rows of real operational data with 9,994 rows of synthetic data to develop robust prediction frameworks. Employing advanced preprocessing techniques such as SMOTE and ADASYN, we implemented five classification models (Decision Tree, k-Nearest Neighbors (kNN), CatBoost, Convolutional Neural Network (CNN), and Long Short-Term Memory (LSTM)) as part of the Reuse Identification Classification Model, which categorizes the likelihood of reusing DSRS. Subsequently, three regression models (Ridge Regression, Lasso Regression, and Random Forest) were applied in the Long-Term Utilization Regression Model to estimate long-term usability based on decay trends and activity levels. Our findings reveal that kNN outperforms other classifiers, achieving an AUC-ROC of 0.987, while Ridge Regression and Random Forest yield nearly perfect R-squared values, demonstrating superior long-term prediction accuracy. This study shows that machine learning has the possibility to improve DSRS management by accurately predicting reuse opportunities and estimating long-term requirements. A combination of real and synthetic data has produced models that aid in providing a more operational and data-driven radioactive waste management scheme. The results are significant for policymakers and other stakeholders in making informed decisions to enhance the sustainability and safety of radioactive waste handling.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114272"},"PeriodicalIF":1.9,"publicationDate":"2025-07-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144549363","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":"Core design and performance of the Westinghouse lead fast reactor with UO2 and MOX configurations","authors":"Nicolas Stauff ,&nbsp;Fausto Franceschini ,&nbsp;Kyle Ramey ,&nbsp;Scott Richards ,&nbsp;Jun Liao ,&nbsp;Martin Jonson ,&nbsp;Paolo Ferroni","doi":"10.1016/j.nucengdes.2025.114266","DOIUrl":"10.1016/j.nucengdes.2025.114266","url":null,"abstract":"<div><div>Westinghouse partnered with Argonne National Laboratory to design, model and optimize UO₂- and MOX-fueled core designs for a medium size (950 MWt) Lead Fast Reactor that was pursued by Westinghouse. Using Argonne’s suite of reactor analysis codes together with Westinghouse fuel cost economic models, thousands of candidate cores were considered to achieve the economics-optimized cores presented in this paper. This optimization process considered detailed reactor physics, fuel performance, transient performance, and fuel economics models. The reactor performance of the resulting optimized UO₂- and MOX-fueled core designs are described and compared in this paper. Both cores show fuel performance and transient behavior that is considered acceptable for the optimization presented herein, while further testing campaigns on material performance in high-temperature liquid lead will be required to confirm acceptability at the operating conditions chosen. A multi-batch strategy was selected for the UO₂ core for best fuel utilization with minimum fuel inventory costs. A single-batch fuel management was instead selected for the MOX core to maximize cycle length and minimize the impact of the longer refueling outage resulting from the higher decay heat of the discharged MOX fuel relative to the discharged UO₂ fuel, requiring a longer cooling time before dry-lift of discharged fuel could take place.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114266"},"PeriodicalIF":1.9,"publicationDate":"2025-07-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144549361","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 magnesium chloride salts on stress corrosion cracking behavior of austenitic stainless steels used in dry storage canister","authors":"Mei-Ya Wang ,&nbsp;Ya-Yun Cheng ,&nbsp;Tsung-Kuang Yeh","doi":"10.1016/j.nucengdes.2025.114282","DOIUrl":"10.1016/j.nucengdes.2025.114282","url":null,"abstract":"<div><div>Austenitic stainless steels, including Types 304, 304L, and 316L stainless steel (SS), are commonly adopted canister materials for dry storage of spent nuclear fuels. When spent fuel storage installations are located near chloride-containing areas, stress corrosion cracking of austenitic stainless steels may take place as dried sea salts deposit on the stressed steel surfaces. The purpose of this study is to evaluate the corrosion behaviors of candidate canister materials with U-bend samples exposed to simulated chloride-containing environments. Various test environments were set up in a glass chamber with periodic spraying of magnesium chloride liquid solutions of three different concentrations at various temperatures and a controlled flow of vapor at a constant relative humidity of 40 % for 1500 h. Prior to the exposure tests, all samples underwent treatments of solution annealing and thermal sensitization.</div><div>After each specific test, surface morphologies and the presence of cracks on the samples were examined via scanning electron microscopy analyses. According to the test results in the presence of magnesium chloride deposits, except for the sensitized 304 SS and 304L SS samples, no cracks longer than 500 μm were observed in the sensitized 316L SS sample at 40 °C. The outcome indicated a better corrosion resistance of 316L SS than those of the other two at this designated temperature. At a higher temperature of 60 °C, 304 SS and 304L SS exhibited more cracks and pits than at 40 °C, and coalescence of pits dominated at an even higher temperature of 80 °C. On the other hand, 316L SS showed mainly pitting corrosion at 40 °C, but pits and cracks were observed at 60 °C. In particular, 316L SS exhibited more and deeper localized cracks originating from pits, while a smaller amount of overall corroded surface area was observed at 80 °C.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114282"},"PeriodicalIF":1.9,"publicationDate":"2025-07-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144536186","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental study on the swelling performance of GMZ24 bentonite buffer material under Fe(II) reducing environment","authors":"Zhang Ming ,&nbsp;Tian Lehan ,&nbsp;Xie Jingli ,&nbsp;Hu Dongke ,&nbsp;Zhou Fangliang ,&nbsp;Zhang Yanfei ,&nbsp;Gu Zhanfei","doi":"10.1016/j.nucengdes.2025.114280","DOIUrl":"10.1016/j.nucengdes.2025.114280","url":null,"abstract":"<div><div>The release of Fe(II) from iron-based waste disposal tanks during the operation of high-level waste (HLW) repository can seriously affect the swelling behaviour of the buffer barrier, with implications for the safety and stability of the repository. Conventional dilatometer are susceptible to Fe(II) corrosion and Fe(II) oxidation by air. In this study, a self-made dilatometer is develop to measure the swelling characteristics of GMZ24 bentonite at different Fe(II) concentrations. The experimental results showed that the stabilisation time of GMZ24 bentonite swelling ratio was shortened from 331.2 h to 32.1 h, and the swelling pressure was shortened from 130.8 h to 22.1 h as the concentration of Fe(II) was increased from 0 to 2.69 mol/L. The final swelling pressure and swelling ratio showed an exponential decrease, with the swelling ratio decreasing from 84.4 % to 30.1 %, and the swelling pressure decreasing from 1050 kPa to 700 kPa. The results of hydrochemical tests and mineralogical analyses showed that the attenuation of the swelling performance of bentonite was mainly attributed to the following two aspects: First the cation exchange between Na⁺ and Fe(II) in bentonite; Second the mineral phase transformation of montmorillonite to berthierine. Based on the ion double layer theory, this study derives relationships describing the variation of GMZ24 bentonite swelling pressure and swelling ratio as a function of Fe(II) concentration. This study provides a scientific basis for predicting the swelling behaviour of GMZ24 bentonite buffer barrier material for HLW repository.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114280"},"PeriodicalIF":1.9,"publicationDate":"2025-07-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144549360","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental investigation on the performance and characteristics of the seawater cooled PRHRS (SWC-PRHRS) under various conditions","authors":"Yong Hwan Yoo ,&nbsp;Wooseong Park ,&nbsp;Hagtae Kim ,&nbsp;Kilsung Kwon ,&nbsp;Yong Hoon Jeong ,&nbsp;Soo Hyoung Kim","doi":"10.1016/j.nucengdes.2025.114279","DOIUrl":"10.1016/j.nucengdes.2025.114279","url":null,"abstract":"<div><div>With the introduction of carbon emission regulations by the International Maritime Organization (IMO), the development of nuclear-powered ships has become increasingly important. Among the key technologies required for such vessels, the design of a passive residual heat removal system (PRHRS) suitable for the marine environment is considered one of the major challenges. The authors previously proposed a conceptual design of a seawater-cooled PRHRS (SWC-PRHRS), utilizing seawater as the ultimate heat sink, and demonstrated its feasibility through both numerical simulations and experimental validation. In this study, the performance and characteristics of the SWC-PRHRS were experimentally investigated under various assumed conditions that may occur in practical applications. These conditions include increased flow resistance within the system, the presence of non-condensable gases such as air, operation using only the seawater heat exchanger for heat removal, inclination of the air heat exchanger, and partial blockage of airflow through selected channels of the air cooler. The study also identified the system’s optimal filling ratio and analyzed the fluid behavior in the connecting pipe between the outlet of the air heat exchanger (AHX) and the inlet of the seawater heat exchanger (SWHX), which plays a critical role in driving head generation and natural circulation. Experimental results showed that the system’s performance was sensitive to flow resistance, while the presence of non-condensable gases had a noticeable effect but remained acceptable under the tested conditions. System performance deteriorated when airflow through some air cooling channels of the cooling tower was obstructed, and SWC-PRHRS could not function using only the seawater heat exchanger. Furthermore, optimal performance was achieved when the filling ratio was within a specific range. Based on prior numerical results, the experiments also confirmed that the internal flow within the connecting pipe between the AHX and the SWHX did not exhibit two-phase behaviors such as mixture or slug flow. Instead, vapor and liquid phases remained clearly separated under gravitational influence.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114279"},"PeriodicalIF":1.9,"publicationDate":"2025-07-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144536187","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":"Material cross-sections real-time inversion from neutron detector data for nuclear reactor digital twin","authors":"Honghang Chi,&nbsp;Jiancheng Chen,&nbsp;Yahui Wang,&nbsp;Yu Ma","doi":"10.1016/j.nucengdes.2025.114296","DOIUrl":"10.1016/j.nucengdes.2025.114296","url":null,"abstract":"<div><div>The Nuclear Reactor Digital Twin (NRDT) has garnered significant attention in recent years. One of the crucial aspects in NRDT is the real-time inversion of nuclear reactor core material cross-sections during reactor operation. In general, an inverse problem is solved by combining multiple iterations of the forward problem with an optimization algorithm. Even though the development of a surrogate model has significantly enhanced the computational efficiency of forward problems, the iteration process still poses a challenge to real-time inversion. To address this problem, this paper presents a real-time inverse problem solver (RIPS). During the offline stage, RIPS establishes a mapping between the sparse neutron detector data and the cross-sections through the reduced-order model and radial basis function. During the online stage, the corresponding cross-section can be calculated directly using the mapping and neutron detector data. Since the RIPS eliminates the multiple iterations of traditional methods, the efficiency of RIPS can be improved by orders of magnitude, and enables real-time online calculation. Three typical numerical benchmarks are tested for verification in this paper, which proves that the maximum relative error of RIPS does not exceed 0.54 % and the average relative error does not exceed 0.1986 %. Furthermore, for each test case, the calculation time of RIPS is within 0.01 s. This work can provide useful suggestions and applications, and further development in cross-section inversion.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114296"},"PeriodicalIF":1.9,"publicationDate":"2025-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144517698","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Review of conceptual design and fundamental research related to the passive residual heat removal system in molten salt reactors","authors":"Shuaiyu Xue ,&nbsp;Chong Zhou ,&nbsp;Pinyan Huang ,&nbsp;Yang Zou","doi":"10.1016/j.nucengdes.2025.114275","DOIUrl":"10.1016/j.nucengdes.2025.114275","url":null,"abstract":"<div><div>The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes such as inherent safety, sustainable development, nuclear non-proliferation, natural resource protection, and economic efficiency. Passive residual heat removal systems for MSRs predominantly encompass Direct Reactor Auxiliary Cooling Systems (DRACS), Salt Discharge Tanks Residual Heat Removal Systems (DTRHRS), and Heat Pipes Residual Heat Removal Systems (HPRHRS). This study introduces an innovative Secondary Side Passive Residual Heat Removal System (SSHRS) for MSRs. The SSHRS employs the primary heat exchanger to dissipate the residual heat from the fuel salt in the primary loop, eliminating the necessity for an additional residual heat removal exchanger and enhancing economic efficiency. The SSHRS approach prevents direct heat transfer from the fuel salt to the environment, mitigates the risk of radioactive material leakage, and bolsters safety. Furthermore, this study also made a horizontal comparison of the advantages and disadvantages of DRACS, DTRHRS, and SSHRS in terms of safety and economy, and discussed the future research directions of passive residual heat removal from molten salt reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114275"},"PeriodicalIF":1.9,"publicationDate":"2025-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144517685","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":"Study on the feasibility of dynamic rod worth measurement using fission chamber detector signal in HPR1000","authors":"Jun Lin ,&nbsp;Xianghui Lu ,&nbsp;Changyou Zhao ,&nbsp;Daifu Wang ,&nbsp;Qingyu Gao ,&nbsp;Yan Zhao ,&nbsp;Xianjun Li ,&nbsp;Mingtao He","doi":"10.1016/j.nucengdes.2025.114255","DOIUrl":"10.1016/j.nucengdes.2025.114255","url":null,"abstract":"<div><div>The control rod worth measurement is one of the most important items during the reactor physical startup tests. Currently the dynamic rod worth measurement (DRWM) based on the signal from non-gamma compensated ion chamber is generally applied on worldwide nuclear power plants, which results in the temporary unavailability of one power range detector and possible insufficiency of zero power physical test range. The fission chamber detector with broader measuring range can transmit effective signal to the digital control system and makes it possible to be applied on the DRWM technique. The calculation model of the HPR1000 ex-core fission chamber detector signal and also the corresponding DRWM spatial-correction factors are built based on OpenMC and PCM and the verification of the calculation model is launched using the measured data from the first cycle of FANGCHENGGANG Unit 3 and Unit 4. The result shows that the calculated and measured variation of current during insertion of control rod are consistent and the DRWM control rod worth using fission chamber detector signals agrees well with that using power range detector signals as well as the theoretical prediction, which proves the feasibility to apply fission chamber detector signals in DRWM.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114255"},"PeriodicalIF":1.9,"publicationDate":"2025-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144523903","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 severe accident diagnosis method based on PCA and DT for a small modular PWR","authors":"Xiaolong Bi,&nbsp;Peiwei Sun,&nbsp;Xinyu Wei","doi":"10.1016/j.nucengdes.2025.114285","DOIUrl":"10.1016/j.nucengdes.2025.114285","url":null,"abstract":"<div><div>Small modular pressurized water reactor (SMPWR) is a new trend in the development of nuclear energy today. SMPWRs usually adopt the integrated arrangement and use many passive safety systems, which have high inherent safety. However, despite all precautions, the severe accident (SA) cannot be completely avoided. It is necessary to conduct rapid and timely diagnosis for these SAs. Different from the traditional SA diagnosis methods based on signals and knowledge, the data-based method is adopted to study SA diagnosis of SMPWR in this study. First, the SA state monitoring method based on principal component analysis (PCA) is proposed to realize rapid and accurate distinction between steady-state and abnormal conditions. Then, the SA classification diagnosis method based on decision tree (DT) is proposed to realize accurate diagnosis of initial events and whether the entry conditions of SA management guideline (SAMG) are met. The key parameters selected through feature selection can provide supplement and reference for the selection of monitoring parameters in SAMG and the determination of the instrument list for instrument availability analysis under SA conditions. The fault diagnosis method proposed in this paper can provide the reference and basis for the SA diagnosis and the design of the operator’s SA handling auxiliary system in the SMPWR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114285"},"PeriodicalIF":1.9,"publicationDate":"2025-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144517686","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":"Plant dynamics analysis of sodium-cooled fast reactors using Flownex code","authors":"Vikram Govindarajan ,&nbsp;Rajendrakumar M. ,&nbsp;Suresh Kumar R. ,&nbsp;Natesan K.","doi":"10.1016/j.nucengdes.2025.114267","DOIUrl":"10.1016/j.nucengdes.2025.114267","url":null,"abstract":"<div><div>Plant dynamics analysis plays an important role in the design and operation of sodium-cooled fast reactors (SFRs). This paper presents the development and validation of a plant dynamic model for an SFR using the Flownex code, a general-purpose thermal-fluid simulation software. A general modeling philosophy is provided for building Flownex models to simulate key components of the SFR, including the core, plenum, pipelines, intermediate heat exchanger (IHX), and pump. A new user-defined script for SFR kinetics calculations has been developed, which enables neutron kinetics calculations based on the point kinetics model with reactivity feedback effects. The procedures for simulating heat sources, sinks, and the inclusion of temperature-dependent sodium fluid properties are discussed in detail. These models and approaches are designed to optimize execution speed while maintaining good accuracy based on practical experience with the code.</div><div>A plant dynamics model is developed for the “Fast Flux Test Facility (FFTF)” reactor using Flownex and is used to simulate the “Loss Of Flow Without Scram (LOFWOS)” test #13. The good agreement observed between the simulation results of various SFR parameters and the experimental data demonstrates the suitability of the Flownex code for advanced plant dynamics studies of SFRs. However, the current system modeling approach has certain limitations, primarily due to the neglect of spatial (multidimensional) effects and the simplified treatment of feedback reactivity components. Potential directions for future improvements are also discussed in this paper.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114267"},"PeriodicalIF":1.9,"publicationDate":"2025-06-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144513791","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}