压水堆不锈钢应力腐蚀开裂行为机理及预测模型

IF 2.1 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Nuclear Engineering and Design Pub Date : 2025-04-01 Epub Date: 2025-02-17 DOI:10.1016/j.nucengdes.2025.113897
Pengfei Gao , Yanhui Li , Zhouyang Bai , Shaoming Ding , Yinan Zhang , Limei Xing , Zhihong Yu
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引用次数: 0

摘要

应力腐蚀开裂(SCC)一直是合金在亚临界水系统中的重要研究课题,也是压水堆(pwr)面临的紧迫问题。由于材料、环境和应力相互作用的复杂性,目前还没有一致的SCC机制和具有力学可拓性的预测模型。本文在回顾不锈钢自裂开裂的典型特征和关键影响因素的基础上,系统总结和比较了较为主流的开裂机理,并从经验和确定性的角度出发,通过实际数据探讨了一系列具有较高识别率的自裂开裂预测模型的适用条件。对于加强读者对SCC关键参数的理解以及各机制的配套使用环境,促进对压水堆SCC行为和累积损伤的准确预测具有重要意义。
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Behavior mechanism and prediction models of stress corrosion cracking of stainless steels in pressurized water reactors
Stress corrosion cracking (SCC) which has long been an essential topic for alloys exposed to subcritical water systems,is an urgent issue for pressurized water reactors (PWRs). Due to the complexity of the interaction of materials, environment, and stress, the SCC mechanism of consensus and the prediction model with mechanic extensibility are unavailable. Based on reviewing the typical characteristics and critical influence factors of SCC of stainless steels, this paper systematically summarizes, compares the more mainstream cracking mechanisms, and from the perspective of empirical and deterministic methods, discusses the applicable conditions of a series of SCC prediction models with high recognition via practical data, which are of great significance for strengthening readers’ understanding of SCC critical parameters and the matching service environment of each SCC mechanism, promoting the accurate prediction of SCC behavior and cumulative damage in the pressurized water reactors.
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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.
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