Evaluation of probabilistic distribution of SCC growth rates obtained under the same test conditions in cold worked stainless steel

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design Pub Date : 2024-11-11 DOI:10.1016/j.nucengdes.2024.113669

Dan Akazawa, Masato Koshiishi, Yasufumi Miura, Kenji Kako

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引用次数: 0

Abstract

Probabilistic fracture mechanics (PFM) is a structural integrity assessment methodology that quantifies failure probability of components in nuclear power plants. In PFM analysis, probabilistic distributions expressed as probabilistic density functions are given to input parameters. This paper discusses the probabilistic distribution of SCC crack growth rates (CGRs) for cold worked Type 316L stainless steel in tests simulating a BWR environment. To assess the probabilistic distribution of these SCC CGRs, the associated data for 40 data of SCC CGRs were obtained from the single heat material under the same test conditions. Both normal and lognormal distributions were in agreement, and the standard deviation was much smaller than previously reported. The PFM results were strongly influenced by the standard deviation of the SCC CGR, suggesting that it is important to consider the SCC CGR probabilistic distribution for a reliable assessment of PFM.

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对冷加工不锈钢在相同试验条件下获得的 SCC 增长率的概率分布进行评估

概率断裂力学（PFM）是一种结构完整性评估方法，用于量化核电站部件的失效概率。在概率断裂力学分析中，输入参数采用概率密度函数表示的概率分布。本文讨论了在模拟 BWR 环境的试验中冷加工 316L 型不锈钢 SCC 裂纹生长率 (CGR) 的概率分布。为了评估这些 SCC CGR 的概率分布，从相同试验条件下的单热材料中获得了 40 个 SCC CGR 的相关数据。正态分布和对数正态分布结果一致，标准偏差比以前报告的要小得多。PFM 结果受 SCC CGR 标准偏差的影响很大，这表明考虑 SCC CGR 的概率分布对于可靠评估 PFM 非常重要。

本文章由计算机程序翻译，如有差异，请以英文原文为准。

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来源期刊

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.