Annals of Nuclear Energy最新文献

Digital twin model generation for crack propagation in steam generator tubes: Methods and implementation 蒸汽发生器管裂纹扩展的数字孪生模型生成：方法与实现

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-07-05 DOI: 10.1016/j.anucene.2025.111697

Yingying Jiang , Hong Xia , Jinming Zhang , Yihu Zhu , Kaige Zhang , Xueying Huang , Wenzhe Yin , Peiqi Jiang

{"title":"Digital twin model generation for crack propagation in steam generator tubes: Methods and implementation","authors":"Yingying Jiang , Hong Xia , Jinming Zhang , Yihu Zhu , Kaige Zhang , Xueying Huang , Wenzhe Yin , Peiqi Jiang","doi":"10.1016/j.anucene.2025.111697","DOIUrl":"10.1016/j.anucene.2025.111697","url":null,"abstract":"<div><div>Crack propagation in the steam generator tubes are critical factor influencing the safe operation of nuclear power plants. Accurate prediction of crack growth behavior is vital for effective equipment health monitoring and maintenance. A fundamental framework and key technologies for the implementation of digital twin technology in nuclear power plants were presented in this study. And a novel crack propagation digital twin model generation technique for steam generator tubes was introduced, which integrates experimental data with extended finite element simulation results. Particle Filter method and reinforcement learning were conducted to evaluate respective strengths and limitations in crack propagation prediction. Both experimental data and extended finite element simulation data were incorporated into the model training process. The results demonstrated that both digital twin model generation techniques have distinct applicability conditions and high accuracy. This study could provide new insights and technological approaches for the health monitoring of steam generator tubes.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111697"},"PeriodicalIF":1.9,"publicationDate":"2025-07-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144563722","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

Machine learning prediction for thermal–hydraulic parameters of semicircular-fin fuel bundle in lead–bismuth fast reactor 铅铋快堆半圆翅片燃料束热水力参数的机器学习预测

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-07-04 DOI: 10.1016/j.anucene.2025.111696

Qian Li , Siwei Cai , Shengcai Zhang , XueChen Liu , Nianmei Zhang , Chen Hu , Xian Zeng , Yulong Mao , Weihua Cai

{"title":"Machine learning prediction for thermal–hydraulic parameters of semicircular-fin fuel bundle in lead–bismuth fast reactor","authors":"Qian Li , Siwei Cai , Shengcai Zhang , XueChen Liu , Nianmei Zhang , Chen Hu , Xian Zeng , Yulong Mao , Weihua Cai","doi":"10.1016/j.anucene.2025.111696","DOIUrl":"10.1016/j.anucene.2025.111696","url":null,"abstract":"<div><div>In this study, a machine learning approach is used to predict the thermal–hydraulic parameters of lead–bismuth eutectic flow in a new-style semicircular-fin rod bundles. With numerical simulation data, a novel micro segment method was used to extract data from 42 sub-channels and 19 fuel rods in the fluid domain and establish a database with 23,313 data points. Four machine learning models are used to predict the <em>h</em> and Δ<em>P</em> by normalizing the input parameters with model hyperparameter optimization. The results show that all of the four machine learning models have good prediction accuracy, with the error of less than 5% and MAPE were all within 1%. The performance of ANN model is better than that of the other three models in predicting new cases. It indicates that ANN model has a high accuracy in predicting thermal parameters under new cases, which verifies the applicability of the machine learning prediction method for multiple cases. This study confirmed the advantages of machine learning in predicting complex regular parameters, and proposed a new method for flow and heat transfer parameters prediction in the development of a subchannel analysis program for semicircular-fin rod bundles.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111696"},"PeriodicalIF":1.9,"publicationDate":"2025-07-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144557041","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

Mathematical modeling and connectivity-enhanced segmentation of void fraction in two-phase flow CT images 两相流CT图像中空隙率的数学建模与连通性增强分割

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-07-04 DOI: 10.1016/j.anucene.2025.111693

Zirou Jiang , Shuo Xu , Jiahao Chang , Jichen Miao , Yuewen Sun

引用次数: 0

Decoupling control for pressurizer pressure and water level in nuclear power plants based on a Proportional–Integral–Derivative Neural Network 基于比例-积分-导数神经网络的核电站稳压器压力与水位解耦控制

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-07-04 DOI: 10.1016/j.anucene.2025.111687

Guofeng Fan, Shengyu Shen, Chunhui Dai

{"title":"Decoupling control for pressurizer pressure and water level in nuclear power plants based on a Proportional–Integral–Derivative Neural Network","authors":"Guofeng Fan, Shengyu Shen, Chunhui Dai","doi":"10.1016/j.anucene.2025.111687","DOIUrl":"10.1016/j.anucene.2025.111687","url":null,"abstract":"<div><div>The pressurizer plays a crucial role in nuclear power plants, ensuring reactor safety by maintaining stable pressure and water levels. It is characterized by nonlinearity, time-varying dynamics, and strong coupling. Conventional single-loop PID control systems often fail to decouple pressure and water level effectively, resulting in suboptimal performance, particularly under large load variations. To address these challenges, this study presents a novel decoupling control system based on a Proportional–Integral–Derivative Neural Network (PIDNN). The PIDNN is employed to mitigate the intense coupling effects between control loops. Furthermore, the Dung Beetle Optimizer (DBO) algorithm is integrated to optimize the initial weights of the PIDNN. A three-region pressurizer model is developed in MATLAB/Simulink for validation. The simulation results demonstrate that the proposed method significantly enhances the control performance and dynamic response speed of the pressurizer’s pressure and water level under typical load rejection transients in nuclear power plants, compared to traditional PID control. This method also exhibits advantages in robustness, adaptability, and engineering applicability, offering an innovative and effective solution for the decoupling control of pressurizers in nuclear power plants.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111687"},"PeriodicalIF":1.9,"publicationDate":"2025-07-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144557042","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

Study on the mechanism of temperature fluctuation induced by coaxial jets in nuclear power systems 核电系统中同轴射流引起的温度波动机理研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-07-03 DOI: 10.1016/j.anucene.2025.111669

Guo-Yan Zhou , Ling-Hao Liu , Xue-Yao Xiong , Xing Luo , Shan-Tung Tu

引用次数: 0

Research on flow and heat transfer characteristics of built-in cooling system in shaft sealed reactor coolant pump 轴封式反应堆冷却剂泵内置冷却系统的流动与传热特性研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-07-03 DOI: 10.1016/j.anucene.2025.111698

Kun Tang , Yizhu Yang , Bin Yin , Xiyuan Cao , Wei Fu , Liang Zhao , Hui Dong

{"title":"Research on flow and heat transfer characteristics of built-in cooling system in shaft sealed reactor coolant pump","authors":"Kun Tang , Yizhu Yang , Bin Yin , Xiyuan Cao , Wei Fu , Liang Zhao , Hui Dong","doi":"10.1016/j.anucene.2025.111698","DOIUrl":"10.1016/j.anucene.2025.111698","url":null,"abstract":"<div><div>The built-in cooling system (BCS) constitutes a critical component of reactor coolant pump (RCP), whose structure is fundamentally determined by the nuclear reactor. Four-bearing mechanical-seal centrifugal pump is widely used in RCP. Operating parameters significantly influence the flow and heat transfer mechanisms in BCS, thereby affecting the service life and operational stability of RCP. Current research on the flow and heat transfer characteristics of BCS lacks comprehensive numerical simulations, which results in insufficient understanding of RCP fundamental operational mechanisms. Therefore, it is necessary to analyze the flow and heat transfer characteristics of BCS. A numerical simulation model established to simulate accurately the multi-passage large-span variable cross-section geometric structural characteristics of BCS in a certain RCP. An experimental platform is constructed for verification, and the flow and heat transfer patterns in BCS under standard operating conditions are obtained. Consequently, the influences of operational parameters on resistance and internal flow distribution are investigated.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111698"},"PeriodicalIF":1.9,"publicationDate":"2025-07-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144535866","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

Accuracy analysis of different numerical methods on thermal stratification prediction of liquid metal 不同数值方法对液态金属热分层预测的精度分析

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-07-03 DOI: 10.1016/j.anucene.2025.111690

Shuai Zhang , Yunqing Bai , Wenbo Li , Ming Jin , Tinyu Li , Mei Huang , Qiaoyan Chen , Fenglei Niu

{"title":"Accuracy analysis of different numerical methods on thermal stratification prediction of liquid metal","authors":"Shuai Zhang , Yunqing Bai , Wenbo Li , Ming Jin , Tinyu Li , Mei Huang , Qiaoyan Chen , Fenglei Niu","doi":"10.1016/j.anucene.2025.111690","DOIUrl":"10.1016/j.anucene.2025.111690","url":null,"abstract":"<div><div>The study of thermal stratification is of significant importance for the safety analysis of pool-type sodium fast reactors, and numerical simulation is an essential tool for investigating thermal stratification phenomena. This work evaluates the effectiveness of a couple of CFD solvers in predicting thermal stratification. It compares the impact of different discretization schemes on solution accuracy and analyzes the reasons for observed discrepancies. Several discretization schemes effectively capture thermal stratification, with the high-resolution scheme showing an average error of 4.2 % and a maximum temperature error of 7.7 K, while the second-order upwind scheme shows an average error of 2.59 % and a maximum temperature error of 6.9 K. Additionally, this study investigates the impact of different order difference schemes on the simulation results of thermal stratification. For conditions where the Richardson number (<em>Ri</em>) is greater than 0.82, the simulation results of the first-order upwind scheme are similar to those of the second-order upwind scheme.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111690"},"PeriodicalIF":1.9,"publicationDate":"2025-07-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144548690","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

Study on flow characteristics and formation mechanism of air-water periodic flooding in two 3 × 3 rod bundle channels 2个3 × 3杆束通道气-水周期性驱流特性及形成机理研究

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-07-03 DOI: 10.1016/j.anucene.2025.111706

Guangyuan Jin , Weilian Li , Jinghu Bai , Rui Wang

引用次数: 0

High-precision eigenvalues of the neutron transport equation by the Carlvik method 用Carlvik方法计算中子输运方程的高精度特征值

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-07-02 DOI: 10.1016/j.anucene.2025.111672

Paolo Saracco , Nicolò Abrate , Sandra Dulla , Piero Ravetto

引用次数: 0

Introducing the SLICE Method for estimating pebble-bed reactor inventories at equilibrium operation with SCALE 介绍了用SCALE估计球床反应器平衡运行库存的SLICE方法

IF 1.9 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-07-02 DOI: 10.1016/j.anucene.2025.111684

Friederike Bostelmann, Steven E. Skutnik, William A. Wieselquist

{"title":"Introducing the SLICE Method for estimating pebble-bed reactor inventories at equilibrium operation with SCALE","authors":"Friederike Bostelmann, Steven E. Skutnik, William A. Wieselquist","doi":"10.1016/j.anucene.2025.111684","DOIUrl":"10.1016/j.anucene.2025.111684","url":null,"abstract":"<div><div>This paper introduces the SCALE Leap-In method for Cores at Equilibrium (SLICE) for estimating pebble-bed reactor equilibrium core isotopic inventories using capabilities in the SCALE code system, requiring only a small computational cluster and a few days of computation. This method uses an iterative approach that relies on (1) a surrogate spectrum model that captures spatial and time-dependent spectral conditions, (2) a multi-pass model that captures the pebble’s evolving nuclide inventory as a function of location and time in the core, and (3) a full-core model that captures the core’s spatial neutron flux distribution.</div><div>The SLICE approach is applied to a generic fluoride salt–cooled high-temperature reactor, demonstrating fuel inventory convergence through nuclide concentration inspection across iterations and comparisons for core realizations with varying discretizations. Results agree within <span><math><mo>∼</mo></math></span>5% with another state-of-the-art code, with differences attributed to input parameter or modeling assumption variations in the equilibrium generation methods.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111684"},"PeriodicalIF":1.9,"publicationDate":"2025-07-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144523822","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