Annals of Nuclear Energy最新文献

{"title":"Analyzing the interaction of magnetism, Joule heating, and entropy generation in an inclined porous media filled with ferrofluid","authors":"Siam Abrar Saad ,&nbsp;Srinivas Reddy Kallem ,&nbsp;Siva Reddy Sheri ,&nbsp;Sumon Saha","doi":"10.1016/j.anucene.2025.111801","DOIUrl":"10.1016/j.anucene.2025.111801","url":null,"abstract":"<div><div>This study presents the interplay of magnetohydrodynamics (MHD), Joule heating, and entropy generation in the natural convection heat transfer of a heat-generating Fe₃O₄–water ferrofluid within an inclined porous square cavity featuring discrete heating and cooling sources. Understanding this complex interaction is vital for optimizing thermal systems in applications such as nuclear reactors, electronic cooling, and process industries, where efficient heat transfer and minimal irreversibility are anticipated. The research employs a numerical approach using the Galerkin finite element method to solve the two-dimensional Navier–Stokes and thermal energy equations under the Darcy–Brinkman–Forchheimer model for porous medium flow. Parametric simulations are conducted across a range of Rayleigh numbers (<span><math><mrow><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>9</mn></mrow></msup><mo>≤</mo><mi>R</mi><mi>a</mi><mo>≤</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>12</mn></mrow></msup></mrow></math></span>), Hartmann numbers (<span><math><mrow><mn>0</mn><mo>≤</mo><mi>H</mi><mi>a</mi><mo>≤</mo><mn>50</mn></mrow></math></span>), inclination angles (<span><math><mrow><mn>0</mn><mo>°</mo><mo>≤</mo><mi>γ</mi><mo>≤</mo><mn>45</mn><mo>°</mo></mrow></math></span>), internal heat generation factors (<span><math><mrow><mn>0</mn><mo>≤</mo><mi>Δ</mi><mo>≤</mo><mn>1</mn></mrow></math></span>), and porous materials (copper powder, brick, and soda-lime silicate). Key findings reveal that thermal performance improves significantly at higher Rayleigh numbers and greater inclination angles (up to 62%) in the ferrofluid filled copper powder porous domain. Conversely, internal heat generation reduces heat transfer efficiency and increases entropy generation. However, the MHD effect exerts negligible influence on the system’s overall efficiency due to suppressed ferrofluid motion in the porous medium. Interestingly, swapping porous medium to soda-lime silicate enhances thermal efficacy (up to 300%) and reduces irreversibilities (up to 0.54%) for <span><math><mrow><mi>R</mi><mi>a</mi><mo>≥</mo><mn>5</mn><mo>.</mo><mn>5</mn><mo>×</mo><mn>1</mn><msup><mrow><mn>0</mn></mrow><mrow><mn>11</mn></mrow></msup></mrow></math></span>. This study uniquely integrates the effect of MHD, Joule heating, internal heat generation, and porous media analyzing thermal efficacy and irreversibilities in an inclined ferrofluid-filled porous chamber with discrete heating to identify optimal configurations, which advances beyond existing literature.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111801"},"PeriodicalIF":2.3,"publicationDate":"2025-08-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144863429","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-based enhancement of integrated process control performance in nuclear reactor operations","authors":"Hui-Yu Hsieh,&nbsp;Thabit Abuqudaira,&nbsp;Pavel Tsvetkov","doi":"10.1016/j.anucene.2025.111765","DOIUrl":"10.1016/j.anucene.2025.111765","url":null,"abstract":"<div><div>This study proposes a hybrid intelligent Proportional Integral Derivative (PID) control approach that dynamically adjusts proportional (K_p), integral (K_i), and derivative (K_d) gains based on the transient severity level to enhance control performance in a nuclear reactor. The control approach is demonstrated through reactivity insertion and primary pump failure transients in a thermal-spectrum molten salt reactor (MSR) operating at 1 MW_th. It utilized deep learning-based predictive models to estimate power peak, time to peak, and power ramp rate, which served as indicators of transient severity and guided real-time control adjustments. To ensure reliability and accuracy, uncertainty quantification (UQ) methods, including deep ensembles, Monte Carlo (MC) dropout, and Bayesian Neural Networks (BNN) last layer, were applied to assess prediction confidence. Results obtained from the predictive models were aligned with results obtained from simulations using the System Dynamics Analysis Tool (SDAT). Only minor uncertainties were observed across all models. By integrating predictive models, the proposed deep learning-based PID controller method outperformed the genetic algorithm-tuned PID controller, particularly in mitigating power fluctuations during pump failure transients. The findings demonstrate the feasibility and effectiveness of using a deep learning-based PID control approach in maintaining reactor stability and improving reactor safety under transient conditions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111765"},"PeriodicalIF":2.3,"publicationDate":"2025-08-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144866387","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 and verification of thermal hydraulic model of the pebble bed high-temperature gas-cooled reactor core analysis code NECP-Panda","authors":"Dongyu Xu ,&nbsp;Yongping Wang ,&nbsp;Yuxuan Wu ,&nbsp;Aolin Zhang ,&nbsp;Liangzhi Cao ,&nbsp;Hongchun Wu ,&nbsp;Yong Luo","doi":"10.1016/j.anucene.2025.111810","DOIUrl":"10.1016/j.anucene.2025.111810","url":null,"abstract":"<div><div>The pebble bed high-temperature gas-cooled reactor (PB-HTGR) has become a research hotspot in recent studies. Calculating the temperature distribution of the pebble bed and within the fuel pebbles is one of the key tasks in the thermal–hydraulic analysis. The VSOP code is used in the design for the HTGR Pebble Bed Module (HTR-PM) of China, in which THERMIX is the thermal–hydraulic calculation module. However, some calculation models of THERMIX are relatively simplified, such as the homogenized approach used in the heat conduction calculation of the fuel pebble scale, neglecting the thermal characteristics of the TRISO particle scale. To facilitate the localization and autonomy of core design code and achieve more accurate simulations, a three-dimensional thermal–hydraulic code for PB-HTGR, named NECP-Panda-TH, has been developed. For helium flow calculation, the code adopts the finite difference method (FDM) to solve the three-dimensional fluid momentum and mass conservation equations, and applies an exponential approximation to obtain fluid temperature distributions. In terms of solid heat conduction calculation, the Nodal Expansion Method (NEM) is applied to solve the solid heat conduction equation of the pebble bed. In addition, a multi-scale model is applied to obtain the accurate temperature distribution of fuel pebbles and even TRISO particles. The verification is then carried out based on HTR-PM, and the results are compared with both the THERMIX code and the commercial software COMSOL Multiphysics. The comparisons indicate that NECP-Panda-TH is capable of the thermal–hydraulic simulations of the HTR-PM.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111810"},"PeriodicalIF":2.3,"publicationDate":"2025-08-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144863427","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":"Group condensation method for the transport calculation of PWR with dispersed-particle fuel","authors":"Xiang Li,&nbsp;Zhouyu Liu,&nbsp;Siyu Yi,&nbsp;Zhao Tian,&nbsp;Shunhong Wang,&nbsp;Longwen Jiang,&nbsp;Liangzhi Cao,&nbsp;Hongchun Wu","doi":"10.1016/j.anucene.2025.111806","DOIUrl":"10.1016/j.anucene.2025.111806","url":null,"abstract":"<div><div>A new group condensation method for pressurized water reactor (PWR) with dispersed-particle fuel is proposed. This method utilizes the neutron current method to evaluate Dancoff correction factors for different pin-cells, after which equivalent 1-D models are built by preserving the Dancoff correction factor. The Hébert model is employed to calculate the energy spectrum of fuel particles and matrix based on the equivalent model to condense the 69-fine-group cross-sections to ∼ 20 groups. This new method can accurately evaluate the double heterogeneity effect which originates from the macro heterogeneity between fuel rods and the micro heterogeneity between particles. Verification calculations for fuel assembly and JRR-3 plate-type reactor show that this group condensation method can cut the CPU time spent on transport calculation by about 60 % while ensuring that the reactivity error is less than 65 pcm and the maximum error in assembly powers is within 1.10 % with a proper broad-group structure.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111806"},"PeriodicalIF":2.3,"publicationDate":"2025-08-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144863428","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":"Beyond EPZs: nuclear emergency management frameworks when using site boundary emergency planning zones, with a case study on high temperature gas reactors","authors":"Luke Lebel,&nbsp;Feng Zhou ,&nbsp;David Hummel","doi":"10.1016/j.anucene.2025.111808","DOIUrl":"10.1016/j.anucene.2025.111808","url":null,"abstract":"<div><div>Adapting emergency management programs to the risk profile of small modular reactors, while ensuring that they still meet necessary defence in depth safety objectives, is a key pillar in the responsible deployment of the technologies. Predicated on the assumption that urgent protective actions would not be required even following a bounding emergency, a <em>site boundary EPZ</em> approach allows for a measurement-driven framework for implementing protective actions. It is a reactive approach that uses dose measurements and operational intervention levels (OILs) to trigger the early (non-urgent) protective actions aimed at averting doses over the longer term, and confirm assumptions about urgent protective actions. A case study based on a high temperature gas reactor (HTGR) safety analysis is performed to show the potential source term and off-site consequences, develop HTGR-specific OILs, and show how different non-urgent protective actions can be triggered with measurements exceeding OIL values.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111808"},"PeriodicalIF":2.3,"publicationDate":"2025-08-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144858126","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 a hybrid neural network framework for CHF prediction in a wire-wrapped rod bundle","authors":"Wei Zhang ,&nbsp;Junsen Fu ,&nbsp;Shuo Chen ,&nbsp;Lijun Yu ,&nbsp;Yao Xiao ,&nbsp;Hanyang Gu","doi":"10.1016/j.anucene.2025.111804","DOIUrl":"10.1016/j.anucene.2025.111804","url":null,"abstract":"<div><div>While predicting critical heat transfer in rod bundles is crucial for reactor safety analysis, most of the existing methods fall short in achieving a balance between accuracy and generalization. Utilizing machine learning algorithms, the hybrid neural network framework was applied to explore prediction methods for critical heat transfer in a wire-wrapped rod bundle. It is concluded that the improvement of the prior model will lead to a better estimation result. Besides, with a larger size, the neural network will certainly improve the estimation performance, but it tends to sacrifice the mean of the data to optimize the variance. The hybrid prediction method, moreover, also has a satisfactory error characteristic at different neural network sizes. Besides, the learning preference of the neural network was also analyzed to elucidate the advantages of hybrid prediction methods. This study proposes and verifies an improved scheme for critical prediction methods which has an outstanding prediction performance over traditional fitting and standalone FCNN. Work facilitates the application of machine learning algorithms in thermal–hydraulic engineering..</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111804"},"PeriodicalIF":2.3,"publicationDate":"2025-08-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144852028","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":"Determining the natural radium isotopes in waters by gamma-ray spectrometry with an HPGe detector","authors":"Marta Lilian Victorino Patricio,&nbsp;Daniel Marcos Bonotto","doi":"10.1016/j.anucene.2025.111814","DOIUrl":"10.1016/j.anucene.2025.111814","url":null,"abstract":"<div><div>This paper describes a novel technique for measuring the activity concentration of the four natural radium isotopes (²²⁶Ra, ²²⁴Ra, ²²³Ra, and ²²⁸Ra) in water samples using gamma ray spectrometry with a high-purity germanium detector <strong>(</strong>HPGe<strong>)</strong>. A ²²⁶Ra standard solution with an activity concentration of 74 Bq/L was prepared, and seven dilutions (ranging from 74 Bq/L to 0.074 Bq/L) were performed to create a calibration curve free of matrix effects. Theoretical curves for the other radium isotopes were derived from this calibration curve by employing the detector’s efficiency. Additionally, isotopic separation was performed using the natural proportions of ²³⁸U and ²³⁵U to determine the specific activity of ²²⁶Ra and ²²³Ra, ensuring more precise and reliable activity results. The developed analytical protocol was applied to quantify radium activity concentration in groundwater samples from different aquifer systems across three Brazilian states.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111814"},"PeriodicalIF":2.3,"publicationDate":"2025-08-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144827801","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":"Numerical modelling of rocking submerged spent fuel baskets under fluid–structure interaction","authors":"Fei Zhao ,&nbsp;Jiaojiao Dong ,&nbsp;Yuchao Wang ,&nbsp;Chen Wang ,&nbsp;Jie Qi ,&nbsp;Wei Qin ,&nbsp;Xiaogang Xu ,&nbsp;Daogang Lu ,&nbsp;Yu Liu","doi":"10.1016/j.anucene.2025.111813","DOIUrl":"10.1016/j.anucene.2025.111813","url":null,"abstract":"<div><div>Rocking structures are widely used in seismic design, including in spent fuel storage and transportation baskets in nuclear power plants. Rocking and overturning analysis of such baskets is essential for evaluating structural safety under seismic loading. To provide radiation shielding and cooling, the baskets are submerged in water, leading to significant fluid–structure interaction (FSI) effects during earthquakes. Previous experiments have demonstrated that FSI strongly influences the rocking amplitude. Finite element models developed for racks are not suitable for simulating the rocking behavior of baskets. This study develops a finite element model that can simulate the rocking motion of an underwater basket and calculate parameters such as the time history of the rocking angle. The model results are validated by comparing them with our previous experimental data. In addition, a parametric study is conducted to investigate the effects of the friction coefficient and aspect ratio on the rocking response. The results indicate that as the width increases and the friction coefficient decreases, the rocking amplitude decreases while the sliding displacement increases. Notably, abrupt changes in rocking amplitude are observed within certain width and friction coefficient ranges. The findings provide valuable insights for designing and optimizing submerged rocking structures.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111813"},"PeriodicalIF":2.3,"publicationDate":"2025-08-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144827800","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":"Numerical simulation of grid-to-rod fretting wear coupling with fuel properties during burnup","authors":"Kangkang Jin ,&nbsp;Shiwei Wang ,&nbsp;Xiaojing Liu ,&nbsp;Tengfei Zhang ,&nbsp;Xiang Chai ,&nbsp;Hui He","doi":"10.1016/j.anucene.2025.111792","DOIUrl":"10.1016/j.anucene.2025.111792","url":null,"abstract":"<div><div>A model of Flow-induced Vibration and Fretting Wear of fuel rod was developed to evaluate the nonlinear vibration and cladding wear of the fuel rod under the coolant excitation force, and fuel properties during burnup for a pressurized water reactor. For a fuel rod equipped with a middle grid, its vibration frequency was predominantly governed by the second-order modal frequency. The influence of initial gap between fuel rod and grid on the wear rates was not linear with maximum value at gap of 0. Fretting wear would produce an extra reduction of 38.94 MPa in the circumferential stress and an increase of 1.64 MPa in the radial stress compared to that by only considering changes in pellet-cladding gap pressure, creep, and thermal expansion over a period of 18 months. These findings suggest that the increase in radial stress constitutes the primary contributor to the elevated risk of fuel failure.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111792"},"PeriodicalIF":2.3,"publicationDate":"2025-08-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144827796","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":"Single-phase flow and heat transfer characteristics of transversely non-uniform heated narrow rectangular channels under typical ocean condition","authors":"Di Zhu,&nbsp;Rulei Sun,&nbsp;Chi Yao,&nbsp;Shixuan Wang,&nbsp;Sichao Tan,&nbsp;Ruifeng Tian","doi":"10.1016/j.anucene.2025.111796","DOIUrl":"10.1016/j.anucene.2025.111796","url":null,"abstract":"<div><div>The self-shielding effect of plate-type fuel elements leads to non-uniform heat generation across the fuel assembly, which significantly influences the temperature distribution on the channel wall surfaces. Additionally, under periodic external forces, the thermal–hydraulic characteristics within the reactor are altered by the additional inertial forces induced by oscillatory accelerations. This study investigates flow and heat transfer in a narrow rectangular channel with transversely non-uniform heating under marine conditions using numerical calculation methods. It is found that under the stationary vertical upward condition, the local Nu number on the transverse cross-section and the wall temperature distribution exhibit significant differences due to variations in transverse heating power distribution. Under ocean conditions, the influence of Coriolis force on local secondary flow and temperature field is predominantly concentrated near the edge regions of the rectangular flow channel. Rolling and pitching motions induce periodic fluctuations in cross-sectional flow characteristics, Nu number, and local wall temperature, with rolling motion causing stronger periodic fluctuations than pitching motion. For the rolling condition, after increasing both the oscillation angle (from 10° to 45°) and the oscillation period (from 10 s to 5 s) of the periodic force field, it is observed that as the oscillatory acceleration increases, the influence of rolling motion on secondary flow intensity progressively intensifies, while the effect of transverse non-uniform heating gradually diminishes.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111796"},"PeriodicalIF":2.3,"publicationDate":"2025-08-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144827802","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}