Prediction of dK/da effects on stress corrosion crack growth rate of irradiated stainless steels based on slip oxidation mechanism

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY Nuclear Engineering and Design Pub Date : 2024-12-01 Epub Date: 2024-09-27 DOI:10.1016/j.nucengdes.2024.113610

Masato Koshiishi

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Abstract

This study developed a predicting method for the crack growth rate (CGR) of stress corrosion cracking (SCC) under varying stress intensity factor (K) conditions for irradiated stainless steel (SS). First, optimization of the Hashimoto-Koshiishi model equation for calculating CGR under varying dK/da conditions was carried out using the Rice, Dugan, and Sham (RDS) equation for the crack tip strain rate. Second, the model input coefficients were set to incorporate the effect of dK/da on SCC CGR to fit the experimental data. Finally, the effect of dK/da on the CGR for 3 dpa irradiated SS of a core shroud was evaluated taking into account the weld residual stress distribution. The model prediction showed that there was little effect on the CGR for the increasing and decreasing K regions in the plate thickness direction of the core shroud under boiling water reactor (BWR) normal water chemistry, but the effect was not negligible for the decreasing K region under BWR hydrogen water chemistry.

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基于滑移氧化机制预测 dK/da 对辐照不锈钢应力腐蚀裂纹生长率的影响

本研究开发了一种预测辐照不锈钢（SS）在不同应力强度因子（K）条件下应力腐蚀开裂（SCC）裂纹生长率（CGR）的方法。首先，利用裂纹尖端应变率的 Rice、Dugan 和 Sham（RDS）方程，优化了用于计算不同 dK/da 条件下 CGR 的 HashimotoKoshiishi 模型方程。其次，设置模型输入系数，将 dK/da 对 SCC CGR 的影响纳入其中，以拟合实验数据。最后，考虑到焊接残余应力分布，评估了 dK/da 对 3 dpa 辐照核心护罩 SS 的 CGR 的影响。模型预测结果表明，在沸水反应堆（BWR）正常水化学条件下，堆芯护罩板厚方向上 K 值增大和减小的区域对 CGR 的影响很小，但在沸水反应堆氢水化学条件下，K 值减小的区域对 CGR 的影响不容忽视。

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.