A Non-Intrusive Reduced-Order model for rapid analysis of thermal stratification in pressurizer surge line

IF 1.9 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY
Chuzhen Peng, Han Zhang, Yongwang Ding, Lixun Liu, Yingjie Wu, Jiong Guo, Fu Li
{"title":"A Non-Intrusive Reduced-Order model for rapid analysis of thermal stratification in pressurizer surge line","authors":"Chuzhen Peng,&nbsp;Han Zhang,&nbsp;Yongwang Ding,&nbsp;Lixun Liu,&nbsp;Yingjie Wu,&nbsp;Jiong Guo,&nbsp;Fu Li","doi":"10.1016/j.nucengdes.2025.113996","DOIUrl":null,"url":null,"abstract":"<div><div>The pressurizer surge line serves to connect the pressurizer and primary circuit in a PWR (Pressurized Water Reactor) system. However, thermal stratification at its junction can induce distortion and stress, potentially damaging the pipes. Computational Fluid Dynamics (CFD) is a common numerical tool, but its time-intensive nature poses challenges for real-time assessment, especially with multiple parameter variations. To address this issue, we developed a rapid analysis method using a non-intrusive reduced-order model. The experimental design is optimized by incorporating the Generalized Subset Design to minimize sample requirements. The reduced-order model of the temperature field was derived using Proper Orthogonal Decomposition. Off-design cases were predicted using Linear, Radial Basis Function, and Radial Basis Function Neural Network interpolation techniques. The resulting temperature field was utilized for stress analysis in the pipe structure. Results indicate that linear interpolation performs best, with a maximum CvRMSE (Coefficient of Variation of the Root Mean Square Error) of 0.038 for temperature and a maximum RMSE(Root Mean Square Error) of −0.02% in predicting the maximum equivalent stress. The Radial Basis Function interpolation is slightly inferior to linear interpolation. It better fits the thermal stratification region but lacks accuracy in identifying its boundaries. This inaccuracy is more sensitive to equivalent stress, resulting in a maximum stress deviation of −0.08% for sharp boundaries. Additionally, the Radial Basis Function Neural Network is unsuitable for current study due to insufficient sample size, resulting in a maximum stress identification deviation of −3.8%. Finally, the POD coefficient is used as a independent variable to interpolate the maximum Von-Mises stress, and the relative errors were controlled within 5%. This study provides a rapid and accurate method to evaluate the temperature distributions and the maximum stresses.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113996"},"PeriodicalIF":1.9000,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":null,"platform":"Semanticscholar","paperid":null,"PeriodicalName":"Nuclear Engineering and Design","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S0029549325001736","RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"","PubModel":"","JCR":"Q1","JCRName":"NUCLEAR SCIENCE & TECHNOLOGY","Score":null,"Total":0}
引用次数: 0

Abstract

The pressurizer surge line serves to connect the pressurizer and primary circuit in a PWR (Pressurized Water Reactor) system. However, thermal stratification at its junction can induce distortion and stress, potentially damaging the pipes. Computational Fluid Dynamics (CFD) is a common numerical tool, but its time-intensive nature poses challenges for real-time assessment, especially with multiple parameter variations. To address this issue, we developed a rapid analysis method using a non-intrusive reduced-order model. The experimental design is optimized by incorporating the Generalized Subset Design to minimize sample requirements. The reduced-order model of the temperature field was derived using Proper Orthogonal Decomposition. Off-design cases were predicted using Linear, Radial Basis Function, and Radial Basis Function Neural Network interpolation techniques. The resulting temperature field was utilized for stress analysis in the pipe structure. Results indicate that linear interpolation performs best, with a maximum CvRMSE (Coefficient of Variation of the Root Mean Square Error) of 0.038 for temperature and a maximum RMSE(Root Mean Square Error) of −0.02% in predicting the maximum equivalent stress. The Radial Basis Function interpolation is slightly inferior to linear interpolation. It better fits the thermal stratification region but lacks accuracy in identifying its boundaries. This inaccuracy is more sensitive to equivalent stress, resulting in a maximum stress deviation of −0.08% for sharp boundaries. Additionally, the Radial Basis Function Neural Network is unsuitable for current study due to insufficient sample size, resulting in a maximum stress identification deviation of −3.8%. Finally, the POD coefficient is used as a independent variable to interpolate the maximum Von-Mises stress, and the relative errors were controlled within 5%. This study provides a rapid and accurate method to evaluate the temperature distributions and the maximum stresses.
稳压器喘振管路热分层快速分析的非侵入式降阶模型
在压水堆(PWR)系统中,稳压器喘振管线用于连接稳压器和一次回路。然而,其连接处的热分层会引起变形和应力,从而潜在地损坏管道。计算流体动力学(CFD)是一种常用的数值计算工具,但其耗时的特性给实时评估带来了挑战,特别是在多个参数变化的情况下。为了解决这个问题,我们开发了一种使用非侵入性降阶模型的快速分析方法。实验设计通过纳入广义子集设计来优化,以最小化样本需求。采用正交分解法,建立了温度场的降阶模型。使用线性、径向基函数和径向基函数神经网络插值技术预测非设计情况。利用所得温度场对管道结构进行应力分析。结果表明,线性插值效果最好,预测温度的最大CvRMSE(均方根误差变异系数)为0.038,预测最大等效应力的最大RMSE(均方根误差)为- 0.02%。径向基函数插值略次于线性插值。该方法较好地拟合了热分层区,但在识别热分层区边界方面缺乏准确性。这种不准确性对等效应力更敏感,导致尖锐边界的最大应力偏差为- 0.08%。此外,由于样本量不足,径向基函数神经网络不适合当前的研究,导致应力识别的最大偏差为−3.8%。最后,以POD系数为自变量插值最大Von-Mises应力,相对误差控制在5%以内。该研究提供了一种快速准确的温度分布和最大应力评估方法。
本文章由计算机程序翻译,如有差异,请以英文原文为准。
求助全文
约1分钟内获得全文 求助全文
来源期刊
Nuclear Engineering and Design
Nuclear Engineering and Design 工程技术-核科学技术
CiteScore
3.40
自引率
11.80%
发文量
377
审稿时长
5 months
期刊介绍: Nuclear Engineering and Design covers the wide range of disciplines involved in the engineering, design, safety and construction of nuclear fission reactors. The Editors welcome papers both on applied and innovative aspects and developments in nuclear science and technology. Fundamentals of Reactor Design include: • Thermal-Hydraulics and Core Physics • Safety Analysis, Risk Assessment (PSA) • Structural and Mechanical Engineering • Materials Science • Fuel Behavior and Design • Structural Plant Design • Engineering of Reactor Components • Experiments Aspects beyond fundamentals of Reactor Design covered: • Accident Mitigation Measures • Reactor Control Systems • Licensing Issues • Safeguard Engineering • Economy of Plants • Reprocessing / Waste Disposal • Applications of Nuclear Energy • Maintenance • Decommissioning Papers on new reactor ideas and developments (Generation IV reactors) such as inherently safe modular HTRs, High Performance LWRs/HWRs and LMFBs/GFR will be considered; Actinide Burners, Accelerator Driven Systems, Energy Amplifiers and other special designs of power and research reactors and their applications are also encouraged.
×
引用
GB/T 7714-2015
复制
MLA
复制
APA
复制
导出至
BibTeX EndNote RefMan NoteFirst NoteExpress
×
提示
您的信息不完整,为了账户安全,请先补充。
现在去补充
×
提示
您因"违规操作"
具体请查看互助需知
我知道了
×
提示
确定
请完成安全验证×
copy
已复制链接
快去分享给好友吧!
我知道了
右上角分享
点击右上角分享
0
联系我们:info@booksci.cn Book学术提供免费学术资源搜索服务,方便国内外学者检索中英文文献。致力于提供最便捷和优质的服务体验。 Copyright © 2023 布克学术 All rights reserved.
京ICP备2023020795号-1
ghs 京公网安备 11010802042870号
Book学术文献互助
Book学术文献互助群
群 号:481959085
Book学术官方微信