Annals of Nuclear Energy最新文献

{"title":"Corrigendum to “Neutronic analysis for withdrawal of TRIGA MARK II reactor control rods using OpenMC program” [Ann. Nucl. Energy 217 (2025) 111330]","authors":"Fajri Prasetya , Zaki Su’ud , Nuri Trianti , Efrizon Umar , Ratna Dewi Syarifah","doi":"10.1016/j.anucene.2025.111483","DOIUrl":"10.1016/j.anucene.2025.111483","url":null,"abstract":"","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111483"},"PeriodicalIF":1.9,"publicationDate":"2025-04-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143890469","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 detailed review on critical flow experiments and models","authors":"Gengyuan Tian ,&nbsp;Yuan Zhou ,&nbsp;Yanping Huang ,&nbsp;Yuan Yuan ,&nbsp;Chengtian Zeng","doi":"10.1016/j.anucene.2025.111441","DOIUrl":"10.1016/j.anucene.2025.111441","url":null,"abstract":"<div><div>Critical flow is a significant phenomenon in reactor safety analysis, which has remained a research focus for several decades. The critical flow rate determines the rate of coolant loss in the reactor and the rate of pressure relief in the primary loop. Owing to the multiplicity of parameters influencing critical flow and the complexity of the flow phenomenon, there exists an incomplete cognition of the critical flow phenomenon at present, and precisely calculating the critical flow rate under various working conditions is extremely challenging. Consequently, this paper presents a comprehensive review of the experimental and theoretical studies on critical flow carried out in various thermodynamic regions. It summarizes the current progress and deficiencies in experimental research and concurrently compiles the experimental research data to offer database support for other researchers engaged in theoretical studies of critical flow. A brief but comprehensive overview of the existing theoretical models and general modeling framework for critical flow phenomena is provided. It is conducive to the further understanding and investigation of the critical flow phenomenon.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111441"},"PeriodicalIF":1.9,"publicationDate":"2025-04-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143822389","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":"Dynamics modeling of molten salt reactor with reduced and expanded representations of delayed neutron precursors","authors":"Thabit Abuqudaira ,&nbsp;Pavel Tsvetkov ,&nbsp;Piyush Sabharwall","doi":"10.1016/j.anucene.2025.111461","DOIUrl":"10.1016/j.anucene.2025.111461","url":null,"abstract":"<div><div>Molten salt reactors (MSRs) present unique challenges in dynamic behavior due to the mobility of their fuel. In these reactors, delayed neutron precursors (DNPs) drift with the fuel circulation through the primary loop. As a result, a fraction of DNPs decays outside the core, effectively reducing the available delayed neutron population for reactivity control. Consequently, precise modeling of the distribution and behavior of DNPs is critical for accurate reactor dynamics simulations. In this study, the System Dynamics Analysis Tool (SDAT) was used to simulate a thermal-spectrum MSR under steady-state conditions and following transients. The effects of using reduced and expanded representations of DNPs with fewer or more groups than the conventional 6-group model were investigated. Their impact on the simulated distribution of precursors in the primary loop, reactivity loss value, and reactor response to transients was analyzed. Simulation results showed that reduced models lead to the loss of the actual DNPs distribution data, resulting in less accurate estimates of reactivity loss. Reactor power predictions using these reduced models showed significant deviations compared to those using the conventional 6-group model in transient simulations. Expanded models offered a more accurate representation of the distribution of DNPs and reactivity loss estimates. Reactor power predictions using expanded models showed minimal deviation from the conventional 6-group model during the simulated transients.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111461"},"PeriodicalIF":1.9,"publicationDate":"2025-04-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143824512","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":"Synthesis and radiation attenuation properties of polymethyl methacrylate/apatite-wollastonite composites for advanced shielding applications","authors":"Norah Salem Alsaiari ,&nbsp;E.O. Echeweozo ,&nbsp;Mine Kirkbinar ,&nbsp;M.S. Al-Buriahi","doi":"10.1016/j.anucene.2025.111466","DOIUrl":"10.1016/j.anucene.2025.111466","url":null,"abstract":"<div><div>This study investigated the microstructure and radiation shielding properties of Polymethyl Methacrylate (PAMM) doped with 5 %, 10 %, 15 %, and 20 % nano-grade pulverized Apatite-Wollastonite (AW) to form a biocompatible material with potential in medical radiation shielding and brachytherapy applications. X-ray diffraction (XRD) and Scanning electron microscopy (SEM) were used to characterize the microstructure and phase composition of the prepared composites. The radiation attenuation properties of the composite were theoretically evaluated for the linear attenuation coefficient (LAC), mass attenuation coefficient (MAC), Effective atomic number (Z_eff), High-value layer (HVL), Mean free path (MFP) and Effective neutron removal cross-sections (Σ_R). The results showed that integrating Apatite-Wollastonite with high radiation attenuation coefficients, into the PMMA matrix enhances the shielding effectiveness of the resultant composite while preserving the necessary mechanical properties. The polymer composite with 20 % AW (P-A-4) exhibited the maximum LAC of 0.038 cm⁻¹ at 15 MeV and Σ_R of 0.03965 cm⁻¹, making it a potential candidate for shielding applications in medical settings, such as around X-ray machines and brachytherapy sources.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111466"},"PeriodicalIF":1.9,"publicationDate":"2025-04-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143822391","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":"Deep learning applications on satellite imagery datasets for nuclear nonproliferation and counter-proliferation","authors":"Jae-Jun Han ,&nbsp;Gayeon Ha ,&nbsp;Youkyung Han ,&nbsp;Changhui Lee ,&nbsp;Hyunjin Lee ,&nbsp;Ahram Song","doi":"10.1016/j.anucene.2025.111443","DOIUrl":"10.1016/j.anucene.2025.111443","url":null,"abstract":"<div><div>This study examined the applicability of deep-learning techniques for extracting artificial structures from high-resolution satellite imagery to support verification processes in nuclear nonproliferation and counter-proliferation efforts. This examination relied on a tailored dataset and an open-source dataset. The tailored dataset was curated using satellite images of well-known nuclear complexes and was further refined to enhance domain relevance. Furthermore, using the attention U-Net model, optimal values of parameters such as batch size were determined to enhance performance. The model was then tested on satellite images of nuclear facilities from various sources, demonstrating effective performance even when applied to distinct and complex environments. To assess the robustness of the model, accuracy evaluations were conducted using both pixel-based and object-based tests. This dual evaluation approach provided a comprehensive analysis of the model, highlighting its practical utility for real-world verification tasks, particularly those related to nuclear activities. Although some false positives were detected, the proposed approach enabled the successful extraction of the majority of structures of interest. This achievement is anticipated to substantially reduce the interpretational workload for analysts and offer a transferable solution for global nuclear monitoring applications.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111443"},"PeriodicalIF":1.9,"publicationDate":"2025-04-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817739","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":"Impact of U-10Mo HALEU Fuel Element Tolerances on the Massachusetts Institute of Technology Reactor safety and operational performance – Neutronics","authors":"Valerio Mascolino ,&nbsp;Sero Yang ,&nbsp;Laura M. Jamison ,&nbsp;Kyle E. Anderson ,&nbsp;Lin-wen Hu ,&nbsp;Dhongik S. Yoon ,&nbsp;John A. Stillman ,&nbsp;Walid Mohamed ,&nbsp;Erik H. Wilson","doi":"10.1016/j.anucene.2025.111307","DOIUrl":"10.1016/j.anucene.2025.111307","url":null,"abstract":"<div><div>The U.S. is coordinating efforts for the conversion of six U.S. high performance research reactors (USHPRR), including one critical assembly from highly enriched uranium (HEU) to low-enriched uranium (LEU). In order to continue the mission of these reactors, including the Massachusetts Institute of Technology Reactor (MITR), and achieve similar performance, high assay low-enriched uranium (HALEU) with a high-density metallic alloy of uranium with 10 wt% molybdenum (U-10Mo) is being considered. Following the preliminary design of the proposed MITR LEU fuel elements using the U-10Mo monolithic alloy, the impact of the fabrication specification was assessed. This work focuses on the analysis of select neutronics characteristics of the MITR LEU core as a function of the variation of the relevant fuel specification parameters (e.g., U-10Mo composition, fuel plate thickness, etc.). A separate article submitted to this journal addresses the impact on the thermal hydraulic performance. The analyses in these works are performed based on an all-LEU conversion management plan identified in previous work, in which only the proposed elements are utilized for achieving the conversion of MITR. The variations of two main neutronics characteristics are assessed as a function of the variability of the specification parameters resulting from the fabrication process: the MITR LEU core reactivity and the fuel cycle length. The main findings of this work show that the MITR core can meet the operational requirements during the LEU transition plan under the limiting fabrication parameter combinations considered. In addition, the analyses show that the dependency of the core neutronics characteristics on the specification parameters is highly linear within the specification tolerances. The rates of variation are reported in detail for each parameter and can serve as a powerful tool for future MITR fuel management in cases such as when HALEU supply is established that may allow additional cycle length or other operational benefits.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111307"},"PeriodicalIF":1.9,"publicationDate":"2025-04-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143822390","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":"Sodium dynamic behavior simulation and analysis under whole large leakage sodium-water reaction accident process","authors":"Xi Bai ,&nbsp;Peiwei Sun ,&nbsp;Gang Luo ,&nbsp;Xinyu Wei","doi":"10.1016/j.anucene.2025.111442","DOIUrl":"10.1016/j.anucene.2025.111442","url":null,"abstract":"<div><div>Under a large leakage sodium-water reaction accident in a sodium-cooled reactor, the whole accident process simulation, including pressure wave propagation (PWP) and long-term (LT) stages, needs to be carried out to understand the accident procedure and role of the protection system. To investigate the consequences of the accident, a whole large leakage sodium-water reaction (WLLSWR) model was derived to evaluate the sodium dynamic behavior. The WLLSWR accident simulation results demonstrated that the secondary loop integrity was ensured by the effective protection system action, regardless of the PWP or the LT stages. The water/steam leakage rate, gas chamber volume of the surge tank, SG rupture disks location, and the feedwater isolation valve time were found to influence the secondary peak pressure at the LT stage. Furthermore, the bursting action of rupture disk 3, the sodium level and the pressure of the surge tank, should be focused at the LT stage.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111442"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817729","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Investigation of added mass coefficient of helical tube bundles in steam generator based on finite element modal analysis","authors":"Xiaodong Chen,&nbsp;Shuang Guo,&nbsp;Wei Tan","doi":"10.1016/j.anucene.2025.111453","DOIUrl":"10.1016/j.anucene.2025.111453","url":null,"abstract":"<div><div>The helical coil once-through steam generator (OTSG) is widely used in the design of integrated pressurized water reactors (IPWRs) due to its compact structure and high heat transfer efficiency. During the operation of OTSG, the helical tube is continuously flushed by the coolant. To prevent flow-induced vibrations (FIV), which can lead to tube wear, damage, or even rupture, FIV calibration is essential throughout the design process. The added mass coefficient, a key parameter in calibration, is closely linked to the prediction of natural frequency and directly impacts the risk assessment of FIV. Therefore, a more comprehensive and in-depth study of the added mass coefficients of helical tube bundles is crucial for the structural optimization and safety analysis of OTSGs. This study investigates the effects of structural parameters, such as pitch ratios and tube bundle arrangements, on the natural vibration characteristics and added mass coefficients of helical tube bundles coiled in the same direction within a liquid environment. Large-scale modal analyses using finite element methods are conducted to explore the trend of the added mass coefficient as a function of varying structural parameters. The results are compared with those of tube bundles coiled in the opposite direction to examine the differences between various coiling configurations. A unified set of recommended curves for the added mass coefficients is proposed to assist engineers in efficiently determining the natural frequency of helical tube bundles with varying pitch ratios in the fluid and avoiding resonance failure. The research outcomes are fundamental to improving FIV analysis methods of helical tube bundles and enhancing the structural integrity of OTSGs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111453"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817738","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":"Simulation of pulsatile flow and heat transfer characteristics of liquid-sodium in annular tight-lattice hexagonal rod bundles","authors":"Ayodeji A. Ala,&nbsp;Zhu Feng,&nbsp;Liu Junyan,&nbsp;Bin Ye","doi":"10.1016/j.anucene.2025.111452","DOIUrl":"10.1016/j.anucene.2025.111452","url":null,"abstract":"<div><div>Liquid metal-cooled reactors are important for the future of nuclear energy production. Some of the proposed reactors’ design and economic objectives necessitate a<!--> <!-->compact, flexible, and reliable core arrangement favourable to a tight lattice, annular fuel rods, and liquid sodium coolant. The thermal–hydraulic and thermo-mechanical characteristics of sodium-cooled annular fuel rods hexagonal tight lattice (P/D = 1.08) core configuration were simulated considering steady and unsteady flow conditions. The extension of the tight lattice to the edge subchannel in the 4 × 4 configuration changed the velocity, temperature, and turbulence intensity distributions compared to the 3 × 3 configuration with the conventional edge subchannel. The flow split ratio between the inner and outer subchannels in the 3 × 3 is ∼ 5.5 % compared to 11 % – 13 % in the 4 × 4 fuel assembly. The flow friction resistance-Reynolds number relationship was consistent with previous findings for square rod assemblies with water as a coolant. Deformation and thermal stresses due to the uneven temperature distribution around the fuel pin peak in the outer clad and at rod positions adjacent to the tight lattice gaps. New correlations were proposed for the transient flow friction resistance-Reynolds number relationships and the flow split ratio in an annular tight-lattice fuel assembly.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111452"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817279","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 semi-implicit Chord Length Sampling method for dispersion fuel analysis","authors":"Zhe Chuan Tan ,&nbsp;Zhi Yuan Feng ,&nbsp;Kok Yue Chan ,&nbsp;Kan Wang","doi":"10.1016/j.anucene.2025.111436","DOIUrl":"10.1016/j.anucene.2025.111436","url":null,"abstract":"<div><div>With the advent of stochastic geometry in nuclear reactors, implicit modeling processes play an increasingly important role in the precise simulation of particle transport in random media. Current implicit modeling methods in RMC utilize Chord Length Sampling (CLS). However, the CLS method experiences significant inaccuracies compared to explicit modeling methods when simulating materials of high scattering and absorbing properties, particularly where absorption interferes with scattering, and is especially prone to errors when simulating non-Markovian stochastic media. A Semi-Implicit CLS (SCLS) method is proposed where previous neutron and particle positions are recorded, while an inclusion sphere is used to maximize the accuracy of the method whilst minimizing the computational expense incurred. The accuracy of the algorithm was then verified against particle distributions generated via explicit modeling methods. The results show that SCLS can significantly improve the accuracy of implicit modeling when simulating non-Markovian dispersion fuel compared to the original CLS method.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111436"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817736","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}