Annals of Nuclear Energy最新文献_第5页

Investigation of added mass coefficient of helical tube bundles in steam generator based on finite element modal analysis 基于有限元模态分析的蒸汽发生器螺旋管束附加质量系数研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-04-10 DOI: 10.1016/j.anucene.2025.111453

Xiaodong Chen, Shuang Guo, Wei Tan

{"title":"Investigation of added mass coefficient of helical tube bundles in steam generator based on finite element modal analysis","authors":"Xiaodong Chen, Shuang Guo, 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}

引用次数: 0

A semi-implicit Chord Length Sampling method for dispersion fuel analysis 用于分散燃料分析的半隐式弦长取样法

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-04-10 DOI: 10.1016/j.anucene.2025.111436

Zhe Chuan Tan , Zhi Yuan Feng , Kok Yue Chan , Kan Wang

引用次数: 0

Simulation of pulsatile flow and heat transfer characteristics of liquid-sodium in annular tight-lattice hexagonal rod bundles

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-04-10 DOI: 10.1016/j.anucene.2025.111452

Ayodeji A. Ala, Zhu Feng, Liu Junyan, Bin Ye

{"title":"Simulation of pulsatile flow and heat transfer characteristics of liquid-sodium in annular tight-lattice hexagonal rod bundles","authors":"Ayodeji A. Ala, Zhu Feng, Liu Junyan, 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 acompact, 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}

引用次数: 0

Numerical investigation on deposition characteristics of ferroferric oxide particles fouling in 2 × 2 petal-shaped fuel rod

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-04-10 DOI: 10.1016/j.anucene.2025.111457

Yu Zhao, Jian Zheng, Zhizhong Tan

引用次数: 0

Sensitivity and uncertainty analysis in pebble-bed reactors: A study using the High-Temperature Code Package (HCP)

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-04-10 DOI: 10.1016/j.anucene.2025.111428

Mahmoud Yaseen, Amr Sadek, Wafaa Osman, Muhammad Altahhan, Xu Wu, Maria Avramova, Kostadin Ivanov

{"title":"Sensitivity and uncertainty analysis in pebble-bed reactors: A study using the High-Temperature Code Package (HCP)","authors":"Mahmoud Yaseen, Amr Sadek, Wafaa Osman, Muhammad Altahhan, Xu Wu, Maria Avramova, Kostadin Ivanov","doi":"10.1016/j.anucene.2025.111428","DOIUrl":"10.1016/j.anucene.2025.111428","url":null,"abstract":"<div><div>The High Temperature Code Package (HCP) provides advanced modeling and simulation tools for High-Temperature Gas-Cooled Reactors (HTGRs). However, despite its capabilities, HCP currently lacks integrated methods for Uncertainty Quantification (UQ) and Sensitivity Analysis (SA). This research aims to implement a statistical framework within HCP by leveraging the DAKOTA toolkit and Python libraries, thereby enabling UQ/SA workflows to evaluate how uncertainties influence the performance of HTGR systems. DAKOTA provides state-of-the-art sampling and analysis methods, which are integrated with HCP’s steady-state and transient multiphysics simulation environments. In this study, a UQ analysis was conducted for both steady-state and transient multiphysics scenarios for a the HTR-200 reactor design. Results demonstrate that the HTR-200 model exhibits robust performance under input uncertainties related to inlet gas temperature, mass flow rate, and reactor power, with variations in Quantities of Interest (QoIs) remaining within expected tolerances. A global SA was the primary focus for a Pressurized Loss of Forced Convection (PLOFC) scenario and a fuel depletion case to further explore the influence of key parameters. An innovative strategy was employed to efficiently compute Sobol sensitivity indices for time-dependent QoIs by using a Gaussian process emulator as a surrogate model for HCP, alongside principal component analysis to reduce the dimensionality of time-series data. The results identified reactor power as the most influential parameter for the PLOFC response, while the outer pebble radius and UO<sub>2</sub> density were found to have the most significant impact on fuel depletion and neutron population.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111428"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817730","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}

引用次数: 0

Investigating thermal and mechanical behavior for two thorium fuel types in a PWR

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-04-10 DOI: 10.1016/j.anucene.2025.111420

Ahmed M. Refaey, Samaa A. Wasfy, Mohga I. Hassan

{"title":"Investigating thermal and mechanical behavior for two thorium fuel types in a PWR","authors":"Ahmed M. Refaey, Samaa A. Wasfy, Mohga I. Hassan","doi":"10.1016/j.anucene.2025.111420","DOIUrl":"10.1016/j.anucene.2025.111420","url":null,"abstract":"<div><div>Thorium, as a nuclear fuel, has been the subject of intensive research for many years. One of the options is the adoption of thorium in currently operated nuclear PWRs. In this work two combinations of thorium fuel are investigated. The first contains thorium with uranium, while the second contains thorium with plutonium. The present research aims at studying the thermal and mechanical behavior of the fuel rod and clad under the steady state conditions. Coupling is applied between MCNP6 and ANSYS-17.2 (FLUENT and Static Structure) codes to obtain temperature and power distribution. An iterative process is associated with the exchange of data between codes to meet the convergence criteria, starting by the thermal–hydraulic ANSYS-FLUENT model to calculate actual temperature distributions of fuel, clad and coolant; these results are used in the MCNP code to determine the axial power distribution in the fuel rod. The value and the location of the maximum thermal stress of the fuel and clad are demonstrated. The results showed that thorium Uranium fuel has a better behavior and thus reducing stress on clad.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111420"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817265","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}

引用次数: 0

Modeling, verification and validation of multiple PWR depletion cycles with DRAGON and PARCS

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-04-10 DOI: 10.1016/j.anucene.2025.111427

B. Meunier, M. Hursin

引用次数: 0

Development of a thermal–hydraulic system code with multi-dimensional modeling for liquid metal pool-type fast reactor and preliminary verification and validation 开发液态金属池式快堆多维建模热液系统代码并进行初步验证和确认

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-04-10 DOI: 10.1016/j.anucene.2025.111440

Jieming Hou , Bo Kuang , Meng Zhao , Shirui Li , Wenjun Hu , Wei Chen

{"title":"Development of a thermal–hydraulic system code with multi-dimensional modeling for liquid metal pool-type fast reactor and preliminary verification and validation","authors":"Jieming Hou , Bo Kuang , Meng Zhao , Shirui Li , Wenjun Hu , Wei Chen","doi":"10.1016/j.anucene.2025.111440","DOIUrl":"10.1016/j.anucene.2025.111440","url":null,"abstract":"<div><div>Liquid-metal pool-type fast reactors exhibit unique transient and steady-state behaviors under operational and accidental conditions due to their complex three-dimensional geometries and flow transport properties. To accurately capture these properties, a system analysis code LIMSAC integrating three-dimensional complex spatial and sub-channel hydrodynamic components, is developed in this paper. The code employs the Jacobi-Free-Newton-Krylov (JFNK) method for solving the field equations and demonstrates excellent robustness in dealing with nonlinear and complex coupled phenomena in multi-physics field problems. The development of the code includes the construction of the governing equations and component models, the numerical discretization scheme and a detailed overview of the solution method. The code’s architecture and modules have been developed and validated through numerical tests and experimental data, showing the ability to accurately simulate the thermal–hydraulic characteristics of liquid metal reactor systems, including transient variations in natural circulation loops and thermal stratification in liquid metal pools. The validation results show that the code is capable of simulating the thermal–hydraulic characteristics of the liquid metal reactor system accurately, including the transient variations of the natural circulation loop and the thermal stratification phenomenon of the liquid metal pool. The simulation results are consistent with experimental data in general, demonstrating the accuracy and reliability of the code. Future work will focus on further analytical and experimental studies to develop multi-dimensional turbulence models, extend the code framework, and add component models to accommodate a wider range of engineering safety analysis needs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":""},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143815109","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}

引用次数: 0

ASTEC core degradation calculations in support of Level-2 Probabilistic Safety Assessment for 1300MWe French reactors: Methodology and preliminary results

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-04-10 DOI: 10.1016/j.anucene.2025.111458

M. Monestier, L. Bellenfant, G. Kioseyian, L. Foucher, L. Laborde

{"title":"ASTEC core degradation calculations in support of Level-2 Probabilistic Safety Assessment for 1300MWe French reactors: Methodology and preliminary results","authors":"M. Monestier, L. Bellenfant, G. Kioseyian, L. Foucher, L. Laborde","doi":"10.1016/j.anucene.2025.111458","DOIUrl":"10.1016/j.anucene.2025.111458","url":null,"abstract":"<div><div>The Institute for Radiation Protection and Nuclear Safety (IRSN) in France has updated its Level-2 Probabilistic Safety Assessment (L2 PSA) for the French 1300MWe Pressurized Water Reactors (PWRs) as part of the decennial safety reevaluation for these specific reactor units. This study was particularly underpinned by computations performed using the IRSN ASTEC V2.2 code. ASTEC, which stands for Accident Source Term Evaluation Code, is the reference integral code employed by IRSN for modeling and predicting the progression of severe accidental sequences. Within this framework, IRSN has conducted a total of 554 simulations of accidental sequences, for both conditions of 100% Nominal Power and of reactor shutdown. These accidental sequences have been defined based on the ASNR results of Level-1 PSA. They encompass the entire spectrum of events starting from initiating event to the point of vessel rupture. Furthermore, these simulations implement state-oriented Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMGs). This paper presents the different calculations carried out, outlines the methodology used to define them and the primary outcomes derived.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111458"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817266","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}

引用次数: 0

Parametric study of the computational model on fuel and fission products characteristics analysis of HTR-PM equilibrium core

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-04-10 DOI: 10.1016/j.anucene.2025.111437

Sohail Ahmad Raza, Yongping Wang, Liangzhi Cao, Yuxuan Wu

{"title":"Parametric study of the computational model on fuel and fission products characteristics analysis of HTR-PM equilibrium core","authors":"Sohail Ahmad Raza, Yongping Wang, Liangzhi Cao, Yuxuan Wu","doi":"10.1016/j.anucene.2025.111437","DOIUrl":"10.1016/j.anucene.2025.111437","url":null,"abstract":"<div><div>Fission products (FPs) release from TRISO-coated particles in Pebble Bed High-Temperature Gas-Cooled reactors (PB-HTGRs) is a critical safety concern, influenced by various input parameters. This study examines the impact of neutron cross-sections, grid resolution, and tracer pebble distribution on fuel behavior, radionuclide inventory, and release rates in the HTR-PM equilibrium core using FIRCS computational framework. A preliminary multi-region strategy has also been proposed to address in-core temperature variations for doppler broadened cross-sections. The results show that FPs concentration and release rates (CRR) generally decrease with increasing cross-section temperatures. The multi-region approach produced CRR values similar to those at high cross-section temperatures. Additionally, cumulative burnup and particle failure fraction (PFF) for average fuel decrease with increasing cross-section temperatures, and the multi-region approaches yield the highest average fuel burnup. Beginning-of-life (BOL) cross-sections significantly underestimate <sup>235</sup>U depletion (by a factor of 2.21) and overestimate discharge burnup (by ∼20 %) compared to burnup-dependent cross-sections. Coarser grids over predict FPs release rates but improve computational efficiency, highlighting a trade-off. Similarly, tracer pebble distribution has a significant effect on release rate variability, but both grid resolution and tracer distribution show minimal sensitivity to discharged fuel actinide concentrations and fuel behavior. This comprehensive analysis highlights the importance of selecting appropriate cross-section libraries, grid resolutions, the proposed multi-region strategy, and tracer pebble distributions for accurate PB–HTGR modeling. The findings provide valuable insights into radionuclide behavior and fuel performance, supporting the development of safer and more optimized PB-HTGR designs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111437"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817267","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}

引用次数: 0