Annals of Nuclear Energy最新文献

{"title":"Malfunction diagnosis based on residence time distribution of radiotracer signals in industrial processes using machine learning techniques","authors":"Mohamed S. El_Tokhy,&nbsp;H. Kasban,&nbsp;Elsayed H. Ali","doi":"10.1016/j.anucene.2024.110976","DOIUrl":"10.1016/j.anucene.2024.110976","url":null,"abstract":"<div><div>Diagnosing problems in industrial processes has always been a complex challenge, frequently obstructed by the complex structure of these systems. The present study presents a robust methodology integrating nuclear radiotracer data with machine learning approaches to improve diagnosis. Radiotracers are used to measure residence time distribution (RTD) as a crucial diagnostic technology. Experiments utilize a Flow Rig System (FRS) to simulate industrial conditions, where a Tc-99 m radiotracer (1 mCi) is injected in Dirac form and monitored with sodium iodide scintillation detectors integrated with an ALTAIX data acquisition system (DAS). Machine learning algorithms are subsequently employed to categorize four RTD signals: normal RTD, small exchange RTD, recirculation RTD, and parallel flow RTD. Identifying these signal kinds is essential for precise system diagnostics. We utilize deep learning via Convolutional Neural Networks (CNNs) for feature extraction and an Artificial Neural Network (ANN) for classification. Additionally, the Binary Tree Growth Algorithm (BTGA) is employed to refine feature selection, improving model efficacy and decreasing processing demands. The deep learning model attains complete identification accuracy while implementing the HP classifier, which enhances processing time and precision. We simulate RTD signals for two scenarios − Perfect Mixers in Series (PMS) and Perfect Mixers with Exchange (PMSEX). We corroborate our results by comparing them with RTD simulation tools, demonstrating significant correlation and concordance. Our Results highlight the efficacy of combining advanced machine learning approaches with new real-time data modelling to enhance diagnostics efficiency and reliability in industrial operations. This method offers a revolutionary technique to improve process optimization and defect identification.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526944","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 comprehensive overview of advancements, applications, and impact of supercritical fluid natural circulation loops","authors":"Santosh Kumar Rai ,&nbsp;Pardeep Kumar ,&nbsp;Mukesh Tiwari ,&nbsp;Vinay Panwar ,&nbsp;Dinesh Kumar ,&nbsp;Vipin Kumar Sharma","doi":"10.1016/j.anucene.2024.110971","DOIUrl":"10.1016/j.anucene.2024.110971","url":null,"abstract":"<div><div>Nowadays, many researches are persistently exploring to comprehend the various characteristic of the supercritical fluid natural circulation loop (SCFNCL) such as use of SCFNCL at normal operating condition as well as a passive system for heat removal from the core, steady state and transient behavior of the loop, heat transfer rate, heat transfer coefficient and optimizing mass flow rate of the loop. In last two decade, a significant research has been seen in the form of analytical, computational and experimental works which highlight the notable use of SCFNCL as an active and passive system in nuclear power plant (NPPs). However, very limited state of arts have been reported based on the loop geometry and their effects, different types of supercritical fluids (SCFs) and applications of the loops. Therefore, steady and transient behaviours of loop in single and parallel channels, thermal–hydraulic (TH) instability, effects of the geometrical and operating parameters on SCFNCL and deterioration of heat transfer (DHT) in SCFNCL are the main emphasis of this review. Performance criteria such as instability, transient, and steady-state requirements, along with methods for containing instability, have been covered. It even emphasizes how crucial it is to validate the numerical codes. Since nuclear reactors use coupled SCFNCL as passive cooling systems, different topologies and combinations of fluids are shown. Very limited experimental studies have been reported in the coupled loop, an initial analysis was conducted and the results demonstrated the effectiveness of the system. The review also demonstrated the need for numerical analysis with using different supercritical fluids and combine with the NPP systems as well as experimental investigations, which can be connected to applications in renewable and sustainable energy.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526849","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":"KPCA-based fault detection and diagnosis model for the chemical and volume control system in nuclear power plants","authors":"Yiqian Sun ,&nbsp;Meiqi Song ,&nbsp;Chunjing Song ,&nbsp;Meng Zhao ,&nbsp;Yanhua Yang","doi":"10.1016/j.anucene.2024.110973","DOIUrl":"10.1016/j.anucene.2024.110973","url":null,"abstract":"<div><div>To study the fault intelligent detection and diagnosis method of nuclear power plant systems and improve the detection and diagnosis effect of internal fault of nuclear power plant Chemical and Volume control System (CVS), this study presents an intelligent <strong>F</strong>ault <strong>D</strong>etection and <strong>D</strong>iagnosis model for the <strong>C</strong>hemical and <strong>V</strong>olume control <strong>S</strong>ystem (FDD-CVS) in nuclear power plants (NPPs). The model is based on failure mode and effects analysis of the CVS system and is implemented by combining kernel principal component analysis (KPCA) with decision tree and support vector machine (SVM). FDD-CVS can rapidly and visually recognize faults in CVS based on independent time-point system parameters, and it is capable of diagnosing fault types and specific fault locations. The model is characterized by clear principles, hierarchical diagnostics, fast diagnostic speed, and visualized results. The model is trained and tested by using the data of the passive nuclear power simulation analyzer. The fault detection rate of FDD-CVS is 96.38%, the false alarm rate is 4.34%, and the average accuracy rate is 98.40%. Overall, the fault monitoring and diagnostic method proposed in this article is innovative and provides valuable references for fault diagnosis research in nuclear power plants.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526940","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 of machine learning approaches for identification of perturbed fuel assemblies in WWER-type nuclear reactors","authors":"A. Kamkar,&nbsp;M. Abbasi","doi":"10.1016/j.anucene.2024.110992","DOIUrl":"10.1016/j.anucene.2024.110992","url":null,"abstract":"<div><div>Enhancing the safety of nuclear power plants relies on the prompt and accurate identification of potential anomalies within the reactor. This paper explores the application of machine learning techniques for the identification and localization of perturbed fuel assemblies in WWER-type reactors. Various machine learning classifiers, spanning the decision tree, random forest, k-nearest neighbors, multilayer perceptron, support vector machine, and 1D-convolutional neural network, are scrutinized for their performance under diverse conditions.</div><div>The methodology encompasses data collection, data preprocessing, hyperparameter tuning, and model evaluation. The necessary dataset is generated using DYNOSIM to simulate all conceivable scenarios related to fuel assembly vibration in a WWER-type reactor. In addition to assessing the models under clear and complete input conditions, a sensitivity analysis is performed to gauge the models’ resilience to detector failures and the introduction of white noise. A comparative analysis of the six machine learning classification models reveals that multilayer perceptron, support vector machine, and 1D-convolutional neural network display the most sturdy classification performance, achieving accuracies of 76.38 %, 70.85 %, and 74.64 %, respectively.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526852","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 investigations and inverse heat transfer analysis to study heat transfer and ablation behaviour of concrete exposed to oxyacetylene flame","authors":"Pedduri Jayakrishna ,&nbsp;Prakash Nanthagopalan ,&nbsp;Arunkumar Sridharan ,&nbsp;Shyamprasad Karagadde ,&nbsp;Anuj Kumar Deo ,&nbsp;Srinivasa Rao ,&nbsp;P.K. Baburajan ,&nbsp;S.V. Prabhu","doi":"10.1016/j.anucene.2024.110991","DOIUrl":"10.1016/j.anucene.2024.110991","url":null,"abstract":"<div><div>The phenomenon of ablation occurring in sacrificial concrete surrounding the nuclear reactor is studied by conducting experiments with oxyacetylene welding. The oxidizing flame produced from the oxyacetylene welding is a potential source of high heat flux and temperature exposed to the ferrosiliceous concrete (contains hematite aggregates) and ordinary concrete (without hematite aggregates). The ablation caused by the high-intensity oxidizing flame is observed to be highly non-uniform. The maximum temperature rise, ablation depths and overall mass loss in ordinary concrete are observed to be higher compared to the ferrosiliceous concrete. The presence of hematite in ferrosiliceous concrete has reduced the heat diffusion, ablation depth and mass of ablated material inside the concrete and hence exhibited better ablation characteristics. A semi-infinite heat transfer model ignoring the effects of chemical reactions are formulated and estimated the approximate interfacial heat flux profiles develop at the interface between the flame and concrete. The outcomes of the study conclude that the ferrosiliceous concrete can withstand the high heat fluxes of the oxidizing flame, and hence, it will withstand the adverse scenario of the interaction of molten corium with concrete walls occurs during the failure of nuclear reactors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526938","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Development of active and passive reactivity control systems for a fast spectrum small modular reactor","authors":"Muhammad Farid Khandaq ,&nbsp;Deokjung Lee","doi":"10.1016/j.anucene.2024.110979","DOIUrl":"10.1016/j.anucene.2024.110979","url":null,"abstract":"<div><div>This paper presents the development of active and passive reactivity control systems implemented in a fast-spectrum small modular reactor (SMR). The design of the control assembly serving as an active system still relies on a conventional rod-typed geometry. To enhance its worth, a thermalizing ZrH_1.6 pin is introduced, which softens the neutrons around the control assembly, leading to increased neutron absorption in the absorber material. Additionally, a control assembly with a partial-length absorber is strategically placed at a certain location to mitigate high peak power levels. Two passive systems have been developed to enhance reactor safety: the gas expansion module (GEM) aimed to mitigate the unprotected loss of flow (ULOF), and the liquid lithium-6 absorber intended to address unprotected transient overpower (UTOP) and unprotected loss of heat sink (ULOSH) scenarios. With a total of 24 GEMs and 18 lithium-6 channels installed, they provide negative reactivity ranging from about 932 to 693 pcm (1.17$ to 1.03$) and 521 to 602 pcm (0.65$ to 0.89$), respectively.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526851","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":"Design of a prismatic CERMET megawatt gas-cooled reactor(PC-MGCR) for deep space exploration","authors":"Yuhan Fan ,&nbsp;Rui Yan ,&nbsp;Guifeng Zhu ,&nbsp;Liang Chen ,&nbsp;Naiwen Zhang ,&nbsp;Yang Zou","doi":"10.1016/j.anucene.2024.110946","DOIUrl":"10.1016/j.anucene.2024.110946","url":null,"abstract":"<div><div>Micro reactor is an ideal choice for future deep space exploration mission due to their advantage of stability, long life and minisize. In this study, a Megawatt Gas-Cooled Reactor concept design using Prismatic CERMET(PC-MGCR) is proposed. It aims to support 1MWe power for both electric thrusters and detection equipment operation in 10 years. The helium-xenon binary gas is simultaneously used for reactor coolant and working medium of the Brayton cycle to simplify the system. The influence of core parameters of fuel pin, core size, radial reflector on reactor miniaturization and neutron economy is investigated by Monte Carlo code OpenMC. The difference of Spectral Shift Absorbers(SSAs), control drum(CD) and control rod(CR) on typical accident critical safety is analyzed. An applicable control rod layout scheme with maximum shutdown depth is determined based on control rod interference and flux distribution. The control rod and the control drum represent a similar variation on the effective multiplication factor, with the maximum differential worth appearing at 60˚ and 17.5 cm, respectively. The core can tolerate the failure of 6 control drums and 1 control rod or all control drums in the dropping accident, which reveals high security. The reactor core optimization result based on miniaturization and falling accident critical safety objectives are further obtained. The parameters of the optimized core are evaluated quantitatively from various aspects. The result shows that PC-MGCR has an extremely hard neutron spectrum and a negative fuel temperature coefficient of −0.258pcm/K with a sensitive control and inherent safety. The power and flux distribution of the active core in different states are obtained. Core power distribution is increasingly flat due to unevenly depletion. The comparison results with existing schemes demonstrates the PC-MGCR exists visible advantages in increasing the fuel inventory, the miniaturization, burnup depth and high-temperature operation. Relevant design and analysis results can provide the reference for the subsequent PC-MGCR.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526939","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":"Nuclear safety characterisation of PBX explosives under low-velocity impact conditions","authors":"Xiaohui Guo,&nbsp;Xingfu Cai,&nbsp;Yonggang Huo,&nbsp;Haowei Wang","doi":"10.1016/j.anucene.2024.110984","DOIUrl":"10.1016/j.anucene.2024.110984","url":null,"abstract":"<div><div>The prevention of accidental ignition in PBX is paramount for ensuring nuclear safety. A computational model, grounded in peridynamics(PD), has been employed to examine the dynamic damage response of PBX under varying conditions. The simulation outcomes suggest that the dynamic damage of PBX displays staged characteristics, attributable to the instability induced by stress wave reflection and damage accumulation. An increase in binder strength results in enhanced energy dissipation within the binder, consequently mitigating the overall damage to the PBX. Changes in microstructural parameters, specifically the specific surface area of the PBX, influence the fracture work required for interfacial debonding, thereby altering the damage characteristics of the PBX. This study enhances our understanding of the dynamic damage mechanisms of explosives by investigating the impact velocity and intrinsic parameters of the explosive.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526853","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":"Fault diagnosis and degree evaluation of steam generator heat transfer tube rupture based on hybrid method","authors":"Yingying Jiang ,&nbsp;Hong Xia ,&nbsp;Zhuoran Zhou ,&nbsp;Wenzhe Yin ,&nbsp;Zhujun Jia ,&nbsp;Xueying Huang ,&nbsp;Jiyu Zhang ,&nbsp;Yihu Zhu","doi":"10.1016/j.anucene.2024.110964","DOIUrl":"10.1016/j.anucene.2024.110964","url":null,"abstract":"<div><div>As critical pressure boundary components in pressurized water reactor, the integrity of steam generator heat transfer tubes is continually challenged by adverse operational environments. Direct monitoring of their condition is unfeasible. A hybrid fault diagnosis method for detecting and evaluating the damage of steam generator heat transfer tubes was proposed by this paper. This method constructed pseudo bond graph model by acquiring in-depth knowledge of steam generators and the causal relationships among its various parameters. To overcome the expert system’s limitation in evaluating the extent of damage, a random forest algorithm was employed for severity assessment. Furthermore, particle swarm optimization-based swarm intelligence method was utilized to optimize the hyperparameters of the random forest model. Performance tests indicated that the proposed method could diagnose damage in heat transfer tubes with high accuracy and promptness and effectively assesses the extent of damage in the heat transfer tubes of steam generators.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142447007","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 increasing the gap ratio on performance and pressure pulsations in a mixed-flow reactor coolant pump","authors":"Lei Liu ,&nbsp;Zuowen Zhong ,&nbsp;Bin Lin ,&nbsp;Junhao Gao ,&nbsp;Chenyu Zhou ,&nbsp;Lijun Liu","doi":"10.1016/j.anucene.2024.110985","DOIUrl":"10.1016/j.anucene.2024.110985","url":null,"abstract":"<div><div>Reactor coolant pumps (RCPs) are important in nuclear reactor coolant systems, with the performance and stability significantly influenced by the small gap between the impeller and diffuser. This study investigates the effects of the gap ratio on the performance and pressure pulsation of CAP1400RCP. Results indicate an upper limit to the gap ratio, beyond which the head-flow rate curve degrades, producing an unacceptable hump. Within the permissible range, however, increasing the gap ratio reduces the non-uniformity of the leading edge (LE) of the diffuser caused by the impeller. Which reduces diffuser losses and enhances the efficiency of RCP. The study further reveals the strongest pressure pulsation at the shroud on the LE of diffuser. By increasing the gap ratio, the amplitudes of both <em>f</em>_BPF and its harmonic frequency significantly decrease, indicating a weaker pressure pulsation. These findings offer valuable quantitative guidance for impeller and diffuser gap design in mixed-flow RCPs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142447004","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}