Annals of Nuclear Energy最新文献

{"title":"A theoretical design procedure for cladding protecting lower head of central measuring shroud from thermal shock damage","authors":"Shu Zheng ,&nbsp;Daogang Lu ,&nbsp;Qiong Cao ,&nbsp;Yuxiong Xue","doi":"10.1016/j.anucene.2025.111338","DOIUrl":"10.1016/j.anucene.2025.111338","url":null,"abstract":"<div><div>The cladding serves as a protective barrier for the central measuring shroud, safeguarding it from thermal shock damage caused by the SCRAM events. To facilitate rapid preliminary design, a theoretical design procedure of the cladding was developed based on thermal, mechanical and creep-fatigue damage theories. Then, the design was performed according to actual operating conditions. It was found that the procedure can reduce design time and computational costs of the design, but needs to be adjusted because of stress concentration, with an adjustment factor of 7.32 for the total thickness design and 22.677 for the layer thickness design. The final design features a total cladding thickness of 6 mm, comprising two layers of 3 mm each. Analysis showed that cladding can mitigate heat conduction from the coolant. Specifically, increasing the cladding thickness from 0 to 6 mm reduced the maximum temperature difference by 49 °C and decreased the maximum stress amplitude by 2.35 × 10⁸ Pa.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111338"},"PeriodicalIF":1.9,"publicationDate":"2025-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143592915","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 coordinate control strategy for load following operation of sodium-cooled fast reactor system","authors":"Qingfeng Jiang ,&nbsp;Jinrong Jin ,&nbsp;Areai Nuerlan ,&nbsp;Yizhe Liu ,&nbsp;Jiashuang Wan ,&nbsp;Pengfei Wang","doi":"10.1016/j.anucene.2025.111318","DOIUrl":"10.1016/j.anucene.2025.111318","url":null,"abstract":"<div><div>The sodium-cooled fast reactor (SFR), as one of the most mature and promising Generation IV reactors, is a potential energy source. Load following operation has flexible regulation performance and is suitable for future development trend of grid peaking operation. The paper proposes a coordinate control strategy for the SFR under load following operation mode. Firstly, a simulation platform for the SFR was developed by adopting MATLAB/Simulink library technology. Secondly, a coordinate control strategy under load following operation mode was developed to coordinate primary sodium loop, secondary sodium loop, and feedwater-steam loop. Finally, typical operational transients such as step and ramp load change transients were simulated to validate the developed coordinate control strategy. The simulation results demonstrate that the reactor power has satisfactory flexibility and the system key parameters have good dynamic characteristics, which ensures expected performances of the SFR.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111318"},"PeriodicalIF":1.9,"publicationDate":"2025-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143593025","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":"SFR core-catcher safety issue: Experimentation and simulation of its ablation by an impinging jet","authors":"A. Avrit ,&nbsp;N. Seiler ,&nbsp;A. Lecoanet ,&nbsp;A. Djermoune ,&nbsp;N. Rimbert ,&nbsp;M. Gradeck","doi":"10.1016/j.anucene.2025.111329","DOIUrl":"10.1016/j.anucene.2025.111329","url":null,"abstract":"<div><div>These coupled experimental and numerical studies support the safety demonstration of dedicated mitigation features in Sodium Fast Reactors (SFR). In order to limit potential power excursion in SFR core in the case of hypothetical severe accident, the implementation of the mitigation devices, called transfer tubes, was considered into actual French SFR core design. More specifically, they connect the core to the lower plenum and enable the early discharge of the corium towards the core-catcher. However, these devices, which effectively reduce the probability of reactivity excursions in the core, emphasize the issue of core-catcher thermomechanical resistance. Contrary to previous reactor core designs, the presence of these tubes could lead to coherent corium jets impinging the core-catcher surface. In the past, numerous studies dedicated to thermal ablation of a solid by a jet have been carried out to characterize the maximum ablated depth. Although these experiments are very valuable at a macroscopic scale, they do not give information about the local physical phenomena governing heat transfer and melting. Keeping this final goal in mind, this study followed a R&amp;D methodology based on an experimental situation using simulant materials enabling direct visualization. In this paper, the ablation of an ice block by an immersed water jet and a free surface jet are investigated in conditions shown to be close to the reactor case. The physical analysis of the heat transfer at jet impingement and in its vicinity is characterized for the immersed jet conditions. Associated to the experimental part, CFD simulations of these tests are performed to validate the calculation methodology (mesh, turbulence models and interface melting…). In the future, simulations will be used in particular to characterize steel/ice transposition biases.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111329"},"PeriodicalIF":1.9,"publicationDate":"2025-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143592917","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":"A study of the application of HFETR and its safety review management principles","authors":"Liang Zhang,&nbsp;Qibing Chen,&nbsp;Mingyan Tong,&nbsp;Guoqiang Liu,&nbsp;Peng Liu,&nbsp;Panqing He","doi":"10.1016/j.anucene.2025.111351","DOIUrl":"10.1016/j.anucene.2025.111351","url":null,"abstract":"<div><div>With the rapid development of national economy and scientific and technological capabilities over the past decades, irradiation test and isotope production of research reactors have grown considerably, and the associated safety risks and controls accompanying the research reactor operations and applications have become increasing critical. With the gradual diversification of irradiation test methods and test materials in research reactors, it is difficult for the relevant regulations and guidance frameworks based on HAF/HAD (nuclear safety laws and regulations of China / nuclear safety guidance of China) to cover all types of applications, creating challenges in safety review and management for research reactor applications in practice. In this study, we have sorted out the relevant laws, regulations and bases of research reactor application safety review, categorized the research reactor applications and historical data of HFETR (High Flux Engineering Test Reactor), analyzed the specific application safety review practices of HFETR, and proposed the management principles for its application safety review. The research findings presented in this paper can provide technical references for safety review processes and management practices pertaining to applications in similar research reactors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111351"},"PeriodicalIF":1.9,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579679","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":"Mapping radionuclide concentrations in the UAE using a Gaussian process Machine learning approach","authors":"Bassam A. Khuwaileh ,&nbsp;Belal Almomani ,&nbsp;Samar El-Sayed ,&nbsp;Rahaf Ajaj ,&nbsp;Yumna Akram","doi":"10.1016/j.anucene.2025.111335","DOIUrl":"10.1016/j.anucene.2025.111335","url":null,"abstract":"<div><div>This study introduces a machine learning approach for mapping soil radionuclide concentrations in the UAE using Gaussian Process (GP) regression. Aimed at enhancing environmental monitoring and public health, the approach utilizes soil samples from across the UAE to create a radiation map (covering Ra-226, Th-232, and K-40 isotopes). GP regression, known for its proficiency in spatial data interpolation, predicts radionuclide levels in areas without direct testing. This method is adept at managing the non-linear spatial complexities inherent in geographic data, offering both a qualitative and quantitative understanding beneficial for decision-making and further sampling strategies. The results reveal the GP model’s capacity to accurately reflect geographic variances in isotopic concentrations, with RMSE values of 13%, 14%, and 22% for Ra-226, Th-232, and K-40, respectively. The model’s success in learning the geographical variations of the concentrations showcases its potential to guide future research by identifying areas of increased radioactivity risk.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111335"},"PeriodicalIF":1.9,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579677","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":"Neutronic analysis for withdrawal of TRIGA MARK II reactor control rods using OpenMC program","authors":"Fajri Prasetya ,&nbsp;Zaki Su’ud ,&nbsp;Nuri Trianti ,&nbsp;Efrizon Umar ,&nbsp;Ratna Dewi Syarifah","doi":"10.1016/j.anucene.2025.111330","DOIUrl":"10.1016/j.anucene.2025.111330","url":null,"abstract":"<div><div>Currently, the TRIGA 2000 Bandung reactor which is a MARK II type has reshuffling some of the fuel elements to maintain its criticality during operation. The reshuffling includes filling and moving fuel which has a different uranium composition for each element. Heterogeneity in the composition of fuel elements has the potential to produce fairly high neutron flux and PPF (Power Peaking Factor) values. This is the main focus that needs to be discussed in this study. The purpose of this study is to conduct a neutronic analysis in the form of calculating the values of k-eff, PPF, neutron flux spectrum, reactions rate, neutron flux distribution, and power density distribution with several control rod withdrawal conditions. Analysis of two types of PPF is also discussed, including APF (Axial Power Peaking Factor) and RPF (Radial Power Peaking Factor). Validation in the form of k-eff and APF parameters of the F7 fuel element has been carried out to see the accuracy of the OpenMC code. The results of the validation of the OpenMC and MCNP neutronic parameters for all control rod withdrawal variations show accurate value agreement with %Δk &lt; 1 %. The results of the k-eff value and excess reactivity indicate that the reactor can operate critically if the control rod withdrawal percentage is &gt;50 %. Increasing the control rod percentage significantly affects the APF and RPF value, which shifts and decreases. At 60 % and 100 % withdrawal, the maximum APF values are 1.332 and 1.245. Meanwhile, the RPF values are 2.042 and 1.898. The withdrawal of control rods has a slight impact on the increase in neutron flux spectrum and the fission-absorption reaction rate in the fuel, particularly at thermal energies. A different condition occurs for the maximum neutron flux in the inner radial ring, which decreases as the control rods are withdrawn. The power density distribution results of the 111 fuel elements show that the reactor core still has heterogeneity. This presents a prospect for future research in designing the TRIGA 2000 Bandung reactor to achieve a more uniform power distribution.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111330"},"PeriodicalIF":1.9,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143593024","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":"Optimal design of marine nuclear power deaerator based on improved particle swarm optimization algorithm","authors":"Jiarui Zhao ,&nbsp;Yanjun Li ,&nbsp;Xu Chen ,&nbsp;Yuan Fu ,&nbsp;Baozhi Sun ,&nbsp;Yuanwei Cao ,&nbsp;Jianxin Shi","doi":"10.1016/j.anucene.2025.111332","DOIUrl":"10.1016/j.anucene.2025.111332","url":null,"abstract":"<div><div>Deaerator is a key equipment in the secondary circuit system, its operating parameters and structural size have a significant impact on the thermal efficiency and arrangement rationality of the system. In this research, a mathematical model of the marine nuclear power deaerator is established, and the influence of thermal and structural parameters on the weight and volume of the deaerator is analyzed. An improved particle swarm optimization algorithm is proposed by introducing Tent chaotic mapping, evolutionary factor and Metropolis criterion, and its performance is verified. Taking the weight and volume minimization of the deaerator as the optimization objectives, the optimal design of the deaerator is carried out using the proposed IPSO algorithm, while satisfying the structural and performance constraints. The optimization results show that the volume and weight of the deaerator can be reduced by 12.979% and 10.213%, respectively, and the feasibility of the optimization design method is proved theoretically.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111332"},"PeriodicalIF":1.9,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579630","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":"Preliminary studies on diversion and misuse for TRISO-fueled heat-pipe-cooled microreactors","authors":"Nelson Snow ,&nbsp;Sweet Shanahan ,&nbsp;Quinton Williams ,&nbsp;Ishita Trivedi ,&nbsp;Emerald Ryan ,&nbsp;Stefano Terlizzi ,&nbsp;Camille Palmer ,&nbsp;Ryan Stewart","doi":"10.1016/j.anucene.2025.111309","DOIUrl":"10.1016/j.anucene.2025.111309","url":null,"abstract":"<div><div>Microreactors are being developed by multiple reactor designers to mass-produce nuclear facilities for worldwide deployment. Including microreactors in a country’s energy portfolio introduces new concerns for international safeguards regarding the number of reactor locations. Microreactors have smaller quantities of nuclear material compared to current light-water reactors, which makes diverting enough material for clandestine purposes difficult for a single reactor. Diversion or misuse of multiple microreactors to obtain a significant quantity introduces a new acquisition pathway. This work examines diversion and misuse scenarios for a realistic heat-pipe-cooled microreactor to understand how these scenarios would effect reactor operations to determine if monitoring key parameters could reduce the burden of inspection on monitoring agencies. We determined misuse caused too drastic of an effect on core operations to be a valid acquisition pathway for the realistic heat-pipe-cooled microreactor. Diversion was explored across varying levels of severity, where potential diversion scenarios could yield one significant quantity of material from between six and eleven microreactors. Through examining the critical control drum angle, excess reactivity, control drum worth, and power distribution changes from nominal were detected over the operational lifetime, which could indicate a divergence from normal operations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111309"},"PeriodicalIF":1.9,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579676","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 a non-intrusive ROM for 5 × 5 rod bundles of PWR using small sample data","authors":"Hao Qian ,&nbsp;Guangliang Chen ,&nbsp;Dong Liu ,&nbsp;Yang Yu ,&nbsp;Yuhang Zhang ,&nbsp;Lixuan Zhang ,&nbsp;Jinchao Li","doi":"10.1016/j.anucene.2025.111347","DOIUrl":"10.1016/j.anucene.2025.111347","url":null,"abstract":"<div><div>This study aims to develop an efficient Reduced-Order Model (ROM) for rod-type fuel assemblies in pressurized water reactors (PWR), essential components of typical PWR cores. This paper integrates Proper Orthogonal Decomposition (POD) with machine learning (ML) for the rapid prediction of the thermal–hydraulic (TH) parameter field in the reactor core, aiding in real-time monitoring and optimization. This paper focuses on validating the model’s feasibility and accuracy for PWR core components operating at 15.5 MPa pressure. To verify the applicability of the POD method in predicting small sample data under actual operational conditions, this study uses actual operational boundaries from nuclear power plants and 121 sets of small sample data as boundary conditions. Extensive Computational Fluid Dynamics (CFD) calculations are conducted to collect data for constructing a snapshot matrix and further developing the ROM. A data-driven approach is employed to explore the relationship between POD coefficients and boundary conditions, achieving rapid prediction of the TH parameter field under any combination of boundary conditions. The study compares the predictive performance of four machine learning algorithms: Radial Basis Function Neural Networks (RBFNNs), K-Nearest Neighbors Regression (KNN), Random Forest Regression (RF), and Gradient Boosting Regression (GBR). Results show that all ML algorithms can construct ROM and predict the TH parameter field with reasonable accuracy for test data. Among them, Random Forest Regression achieves the highest precision, with an average Mean Square Error (MSE) of approximately 3.24 × 10⁻⁴ and a coefficient of determination (R²) of around 0.96. Furthermore, the average prediction time for the detailed three-dimensional parameter field per component is only 15.0 s.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111347"},"PeriodicalIF":1.9,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579678","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":"Parametric investigation of mercury cell electrolysis for isotopic separation of Li using coupled numerical methods","authors":"J. Ahmadi ,&nbsp;M. Sharifi ,&nbsp;M. Aghaie ,&nbsp;S. Yousefi-Nasab ,&nbsp;A. Arzhangmehr","doi":"10.1016/j.anucene.2025.111336","DOIUrl":"10.1016/j.anucene.2025.111336","url":null,"abstract":"<div><div>Fusion reactors will be one of the major resources of future energy. The fuel required for these reactors is deuterium–tritium (D-T). Tritium, which is not naturally available, is produced by the collision of neutrons with ⁶Li in the environment of these reactors. Various compounds, including lithium hydroxide (LiOH) solution, can be used to extract ⁶Li. The natural abundance of ⁶Li is approximately 7.5%, which is insufficient for producing tritium used in fusion reactors and necessitates further enrichment. Various methods, including electrolysis, can be employed to increase ⁶Li enrichment. In this research, the electrolysis exchange (ELEX) method has been used to extract ⁶Li from LiOH with a favorable separation factor. The simulation of this process using the design of a mercury cell has been conducted through Comsol 6.1 software. This geometry (mercury cell) is a sloping space for continuous movement of mercury. Mercury, which is actually the cathode in this system, absorbs lithium isotopes, especially ⁶Li, due to the mobility of this isotope compared to ⁷Li, and increases the percentage of ⁶Li abundance at the output of the system (LiHg). Electrochemical physics, involving the solution of Nernst-Planck equations, and turbulent flow physics, governed by the Navier-Stokes equations, have been employed to carry out the electrolysis process. It should be noted that, given the significance of diffusion coefficient values in the process of isotope separation by electrolysis, these values have also been determined through molecular dynamics simulations. Finally, and based on the parametric study that has been conducted on all the parameters affecting the electrolysis process, a separation coefficient of about 1.05 has been obtained. This separation coefficient, which has also been compared with experimental references, has been stabilized after approximately 3 h of operation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111336"},"PeriodicalIF":1.9,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579680","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}