Annals of Nuclear Energy最新文献

{"title":"Impact of U-10Mo HALEU fuel element tolerances on the Massachusetts Institute of Technology reactor safety and operational performance – Thermal hydraulics","authors":"Sero Yang ,&nbsp;Valerio Mascolino ,&nbsp;Laura M. Jamison ,&nbsp;Kyle E. Anderson ,&nbsp;Lin-wen Hu ,&nbsp;Dhongik S. Yoon ,&nbsp;John A. Stillman ,&nbsp;Walid M. Mohamed ,&nbsp;Erik H. Wilson","doi":"10.1016/j.anucene.2025.111376","DOIUrl":"10.1016/j.anucene.2025.111376","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 facility from highly enriched uranium (HEU) to low-enriched uranium (LEU). In order to continue the mission of these reactors, including the Massachusetts of 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 evaluated. The impact of the fabrication specification and tolerances was assessed following the preliminary design of the MITR LEU fuel elements using the U-10Mo monolithic alloy. This research focuses on the analysis of fabrication specification impact on thermal hydraulics (TH) characteristic of the MITR LEU core as a function of the variation of the relevant fuel specification parameters (e.g., coolant channel gap thickness, fuel plate thickness, etc.). The analyses are performed based on an all-fresh LEU fuel conversion plan identified in a preliminary safety analysis report submitted to the Nuclear Regulatory Commission. The reactor power margin to the onset of nucleate boiling (ONB) is assessed under the limiting safety system settings (LSSS), where a scram occurs, to ensure there is sufficient margin to the reactor safety limit, which is defined by the onset of flow instability that occurs after the ONB. The best estimate plus uncertainty approach is employed to analyze this TH characteristic, which yields realistic results while maintaining adequate conservatism, utilizing a statistical uncertainty propagation method with the STAT7 code. The TH characteristic is analyzed as a function of the variability of the specification parameters resulting from the fabrication process. The main findings of this study show that the MITR core can meet the TH safety and operational requirements at the all-LEU initial core startup (cycle 1), selected transition cycles (most reactive cycle and most limiting cycle: cycle 3 and 5, respectively) and equilibrium (cycle 14) cores under all limiting fabrication parameter combinations considered. In addition, the analyses show that the dependency of the core power margin to ONB on those specification parameters that have the most direct impact on TH performance is non-linear but monotonically decreasing within the specification tolerances. The third order polynomial fit curves are reported in detail for selected limiting cases 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 111376"},"PeriodicalIF":1.9,"publicationDate":"2025-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143825795","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 of saturated flow boiling of water in vertical helically coiled tubes","authors":"Yuqing Su,&nbsp;Xiaowei Li,&nbsp;Xinxin Wu","doi":"10.1016/j.anucene.2025.111464","DOIUrl":"10.1016/j.anucene.2025.111464","url":null,"abstract":"<div><div>Helically coiled tubes are used in the High Temperature Gas-cooled Reactor (HTGR) Once Through Steam Generator (OTSG). However, due to centrifugal forces and secondary flows, heat transfer characteristics inside these tubes may differ from that in straight tubes. A clear understanding of the heat transfer inside helically coiled tubes is essential for the design and operation of the OTSG. In this study, we experimentally investigated saturated flow boiling in helically coiled tubes with a large curvature ratio (<em>δ</em> = 0.109). The experimental parameters cover a broad range. The system pressure is from 3.5 to 7 MPa, mass flux is from 300 to 1100 kg/(m²·s) and heat flux is from 50 to 600 kW/m². Results show that the inner wall temperature distribution is uneven, with the highest temperature on the inner side and the lowest on the outer side. Increasing heat flux enhances the saturated flow boiling heat transfer coefficient. At low steam quality, the heat transfer coefficient is not significantly affected by mass flux variations. However, at higher steam quality, increasing mass flux improves heat transfer. An increase in system pressure enhances the heat transfer coefficient at lower steam qualities but reduces it at higher steam qualities. Six correlations for the saturated flow boiling were evaluated, with the Gungor-Winterton correlation originally developed for straight tubes showing accurate predictions for heat transfer coefficients in helically coiled tubes (MAPE is 12.72 %, RMS is15.36 %). This indicates that even in helically coiled tubes with a large curvature ratio, no significant difference is observed in the saturated flow boiling heat transfer coefficient compared to straight tubes.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111464"},"PeriodicalIF":1.9,"publicationDate":"2025-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143825794","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":"Parameter identification of fluid field based on CFD reduced-order model and 3D-Var data assimilation","authors":"Chuqiao Dai ,&nbsp;Di Yang ,&nbsp;Chunyu Zhang ,&nbsp;Helin Gong","doi":"10.1016/j.anucene.2025.111459","DOIUrl":"10.1016/j.anucene.2025.111459","url":null,"abstract":"<div><div>Data assimilation (DA) significantly improves the accuracy of field state and parameter estimation by merging experimental data with predictions from high-fidelity numerical models. However, despite their precision and high resolution, the substantial computational cost associated with these numerical models often hinders their practical application in DA. To overcome this challenge, this study presents a novel three-dimensional variational (3D-Var) DA framework that leverages a reduced-order model (ROM) for boundary parameter estimation in computational fluid dynamics (CFD) models. The framework utilizes Proper Orthogonal Decomposition (POD)-Galerkin projection to construct the ROM, enabling near real-time solutions and significantly enhancing computational efficiency. Furthermore, a nonlinear observation operator is developed within the reduced basis space, which directly connects parameters with observational data. This approach eliminates the necessity for full-state reconstruction, thereby further streamlining the computational process. Benchmark results indicate that the proposed method achieves high accuracy and robustness, offering optimal background information for subsequent state estimation. This advancement not only reduces computational overhead but also maintains the integrity and reliability of the estimations, making it a promising tool for real-time applications in complex fluid dynamics scenarios.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111459"},"PeriodicalIF":1.9,"publicationDate":"2025-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143825796","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-analytical method for modelling station blackout transients in liquid metal-cooled reactors","authors":"Janne Wallenius,&nbsp;Fredrik Dehlin","doi":"10.1016/j.anucene.2025.111414","DOIUrl":"10.1016/j.anucene.2025.111414","url":null,"abstract":"<div><div>A semi-analytical method for modelling station blackout performance in liquid metal reactors is developed, permitting to identify key factors determining peak temperatures during the transient, and hence to design associated passive safety systems. It is shown that integrity of the fuel cladding during this transient can be ensured by adequate dimensioning of coolant channels, the primary system and the vessel air cooling circuit. These dimensions are determined using algebraic equations and postulated values for a minimum/maximum permissible Reynolds number, dimensionless parameters for the fuel cladding tube geometry and heat sink elevation, a guard vessel height, the nominal core power, permitted temperature gradients in the vessel air cooling system and the air cooling system chimney height. The model suggests that the required coolant volume is a rapidly growing function of core power, and that this volume needs to be 40% larger in a sodium-cooled reactor than in a lead-cooled reactor.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111414"},"PeriodicalIF":1.9,"publicationDate":"2025-04-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143825793","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}