Environmentally assisted fatigue design model of thermally aged cast austenitic stainless steel in high-temperature pressurized water

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design Pub Date : 2025-10-10 DOI:10.1016/j.nucengdes.2025.114504

Yufei Qiao , Hui Zheng , Jibo Tan , Shuangliang Yang , Ziyu Zhang , Jie Li , Xinqiang Wu , Wei Ke

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Abstract

Fatigue tests of Z3CN20.09M CASS were carried out in high-temperature pressurized water. The fatigue life of Z3CN20.09M CASS decreased with increasing thermal aging time (0 ∼ 15000 h at 400 °C), while it slightly affected by the dissolved oxygen (＜5 ppb and 500 ppb). Based on the present results and fatigue data from our previous work, a modified Institute of Metal Research (M–IMR) environmental fatigue model considering thermal aging factors on the environmental fatigue correction factor (F_en) was developed. Compared with Argonne National Laboratory (ANL) and IMR models, the M–IMR model was more accurate in the fatigue life prediction of thermally aged Z3CN20.09M CASSs. The M–IMR model can accurately predict the fatigue life of CASSs in high-temperature pressurized water after thermal aging at temperatures below 450 °C, but not after thermal aging at temperatures above 450 °C, which may be related to microstructure differences caused by thermal aging and the change in the thermal aging mechanism at different temperatures.

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高温加压水中热时效铸造奥氏体不锈钢环境辅助疲劳设计模型

对z3cn20.09 9m CASS进行了高温加压水中疲劳试验。Z3CN20.09M CASS的疲劳寿命随着热老化时间的增加而降低（400 ℃时为0 ~ 15000 h），而溶解氧对其影响较小（<5 ppb和500 ppb）。在此基础上，结合前人工作的疲劳数据，建立了考虑热老化因素对环境疲劳修正系数（Fen）的修正金属研究所（M-IMR）环境疲劳模型。与ANL模型和IMR模型相比，M-IMR模型对热时效Z3CN20.09M cass的疲劳寿命预测更为准确。M-IMR模型能较准确地预测温度低于450 °C的热老化后CASSs在高温加压水中的疲劳寿命，而温度高于450 °C的热老化后则不能预测疲劳寿命，这可能与热老化引起的显微组织差异以及不同温度下热老化机理的变化有关。

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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.