Annals of Nuclear Energy最新文献_第10页

Numerical investigation on deposition characteristics of ferroferric oxide particles fouling in 2 × 2 petal-shaped fuel rod 2 × 2花瓣状燃料棒中氧化铁铁颗粒沉积特性的数值研究

IF 1.9 3区工程技术

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

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) 球床反应器的敏感性和不确定性分析——基于高温代码包（HCP）的研究

IF 1.9 3区工程技术

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

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 研究压水堆中两种钍燃料的热学和力学行为

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 用DRAGON和PARCS对多个压水堆耗尽循环进行建模、验证和验证

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 支持法国1300MWe反应堆2级概率安全评估的ASTEC堆芯退化计算：方法和初步结果

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 HTR-PM平衡堆芯燃料及裂变产物特性分析计算模型的参数化研究

IF 1.9 3区工程技术

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

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

Investigation of coarse mesh acceleration methods for the SN nodal method in unstructured geometries 非结构几何中SN节点法的粗网格加速方法研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-04-09 DOI: 10.1016/j.anucene.2025.111451

Haoxiang Xu , Youqi Zheng , Hongchun Wu , Bowen Xiao

{"title":"Investigation of coarse mesh acceleration methods for the SN nodal method in unstructured geometries","authors":"Haoxiang Xu , Youqi Zheng , Hongchun Wu , Bowen Xiao","doi":"10.1016/j.anucene.2025.111451","DOIUrl":"10.1016/j.anucene.2025.111451","url":null,"abstract":"<div><div>The S<sub>N</sub> nodal method with unstructured nodes is an effective approach for modeling the complicated geometries in solving the neutron transport equation. However, it hits an efficiency bottleneck when triangular nodes are adopted in the modeling. Against this backdrop, this study investigated acceleration methods for the S<sub>N</sub> nodal method based on unstructured coarse meshes to address the efficiency problem. To achieve this, the study first proposed a coarse mesh generation algorithm from arbitrary triangular meshes. Then, various CMFD schemes, including pCMFD, odCMFD, and gCMFD, were developed. The proposed method can process regular triangular meshes in structured geometries of hexagonal and rectangle assemblies, as well as arbitrary triangular meshes generated in unstructured geometries using the Delaunay triangulation method. A set of eigenvalue problems with various mesh counts and geometry types was selected to verify the accuracy and evaluate the performance of different acceleration schemes. Results indicated that an acceleration ratio of up to 2–3 can be achieved for different conditions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"219 ","pages":"Article 111451"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143800222","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

Thermal-hydraulics performance and stability two-phase flow using Al2O3 nanofluids in an open natural circulation loop 在开放的自然循环回路中使用Al2O3纳米流体的热工性能和稳定性

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-04-08 DOI: 10.1016/j.anucene.2025.111424

Dwi Yuliaji , Roy Waluyo , Gatot Eka Pramono , Muhammad Ganjar Putra , Nur Rochman Budiyanto , Shendy Akbar Mariadi , Sunandi Kharisma , Ryan Oktaviandi , Deendarlianto , Indarto , Mulya Juarsa

{"title":"Thermal-hydraulics performance and stability two-phase flow using Al2O3 nanofluids in an open natural circulation loop","authors":"Dwi Yuliaji , Roy Waluyo , Gatot Eka Pramono , Muhammad Ganjar Putra , Nur Rochman Budiyanto , Shendy Akbar Mariadi , Sunandi Kharisma , Ryan Oktaviandi , Deendarlianto , Indarto , Mulya Juarsa","doi":"10.1016/j.anucene.2025.111424","DOIUrl":"10.1016/j.anucene.2025.111424","url":null,"abstract":"<div><div>The thermal-hydraulics performance and stability flow using Al<sub>2</sub>O<sub>3</sub> nanofluids in an open natural circulation loop has been investigated. Experiments were conducted with a gradual increase in heating power from 880 <span><math><mi>W</mi></math></span> up to 1350 <span><math><mi>W</mi></math></span>. The working fluid used Al<sub>2</sub>O<sub>3</sub> with variations of 0.025 <span><math><mrow><mi>wt</mi><mo>%</mo></mrow></math></span>, 0.05 <span><math><mrow><mi>wt</mi><mo>%</mo></mrow></math></span>, and 0.1 <span><math><mrow><mi>wt</mi><mo>%</mo></mrow></math></span>. The two-phase flow pattern was observed using 1 fps range macro photos. PSD (power spectral density) and DWT (discrete wavelet transform) signal processing were used to analyze the thermal-hydraulics performance and flow stability that occurred during the experiments. The result shows that there are three types of oscillations were found in the observations based on heating power, intermittent oscillation with expulsion-refill-incubation stages, sinusoidal oscillation with expulsion and refill stages, and high subcooling stable flow circulation with only one continuous refill stage. PSD and DWT analysis provided solid agreement between the temperature signal and thermal-hydraulics performance and flow stability.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":""},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791319","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

Source term uncertainty analysis of filtered containment venting scenarios in Nordic BWR 北欧沸水堆过滤容器排放方案源项不确定性分析

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-04-08 DOI: 10.1016/j.anucene.2025.111406

Sergey Galushin , Govatsa Acharya , Dmitry Grishchenko , Pavel Kudinov

引用次数: 0