冷加工奥氏体不锈钢在原始环境下应力腐蚀开裂十字试件的力学场试验与数值分析

IF 2.8 2区 工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY Journal of Nuclear Materials Pub Date : 2023-08-15 DOI:10.1016/j.jnucmat.2023.154478
Qi Huang , Yann Charles , Marc Maisonneuve , Cécilie Duhamel , Catherine Guerre , Monique Gasperini , Jérôme Crepin
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引用次数: 0

摘要

采用允许顺序加载的十字形试样,在模拟原生水中对冷加工奥氏体不锈钢进行了应力腐蚀开裂(SCC)试验。在SCC试验后,用扫描电镜研究了裂纹密度和位置。为了分析开裂区域的力学场,对整个机械加载过程进行了有限元模拟,包括应变路径和温度变化。结合各向同性-运动硬化作为本构方程,并通过在室温和340℃下进行的拉伸试验进行了识别。通过将数值应变场与在空气中进行的具有代表性的“非原位”试验中通过数字图像相关获得的实验测量结果进行比较,得到了该模型的部分验证。讨论了试验过程中应变和应力场的变化与SCC试验结束时观察到的裂缝网络的关系。
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Experimental and numerical analysis of mechanical fields on cross-shaped specimens for stress corrosion cracking of cold-worked austenitic stainless steels exposed to primary environment

A stress corrosion cracking (SCC) test was performed on a cold-worked austenitic stainless steel in simulated primary water, using a cross-shaped specimen permitting sequential loading. Crack density and location were investigated by scanning electron microscopy after the SCC test. To analyse the mechanical fields in the cracking areas, finite element simulations of the whole mechanical loading were conducted, involving both strain-path and temperature changes. Combined isotropic-kinematic hardening was used as constitutive equation and identified with tensile tests performed at room temperature and at 340 °C. Partial validation of the model was obtained by comparison of numerical strain fields with experimental measurements obtained by digital image correlation performed on a representative “ex-situ” test in air. Variations of the strain and stress fields during this test were discussed in relation with the cracking network observed at the end of the SCC test.

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来源期刊
Journal of Nuclear Materials
Journal of Nuclear Materials 工程技术-材料科学:综合
CiteScore
5.70
自引率
25.80%
发文量
601
审稿时长
63 days
期刊介绍: The Journal of Nuclear Materials publishes high quality papers in materials research for nuclear applications, primarily fission reactors, fusion reactors, and similar environments including radiation areas of charged particle accelerators. Both original research and critical review papers covering experimental, theoretical, and computational aspects of either fundamental or applied nature are welcome. The breadth of the field is such that a wide range of processes and properties in the field of materials science and engineering is of interest to the readership, spanning atom-scale processes, microstructures, thermodynamics, mechanical properties, physical properties, and corrosion, for example. Topics covered by JNM Fission reactor materials, including fuels, cladding, core structures, pressure vessels, coolant interactions with materials, moderator and control components, fission product behavior. Materials aspects of the entire fuel cycle. Materials aspects of the actinides and their compounds. Performance of nuclear waste materials; materials aspects of the immobilization of wastes. Fusion reactor materials, including first walls, blankets, insulators and magnets. Neutron and charged particle radiation effects in materials, including defects, transmutations, microstructures, phase changes and macroscopic properties. Interaction of plasmas, ion beams, electron beams and electromagnetic radiation with materials relevant to nuclear systems.
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