Investigation of stress corrosion crack propagation characteristics and life prediction for thick-walled double U-groove pipe welds in PWR steam generator

IF 1.9 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY
Baoyin Zhu , Zheng He , Lu Zhang , Shuitao Gu , Xiao Jin , Dungui Zuo , Gongye Zhang
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引用次数: 0

Abstract

Stress corrosion cracking (SCC) is a critical concern in evaluating the structural integrity of pressurized water reactor (PWR) primary pressure boundaries. Material mismatch and welding-induced residual stresses introduce significant challenges in predicting crack propagation paths and service-induced defect lifetimes. This study examines the influence of welding residual stress on SCC in a thick-walled double U-groove pipe dissimilar steel welded joint of a PWR steam generator (SG), and assesses the propagation life of SCC from an engineering standpoint. First, the distribution and stress state of residual stresses in complex SG dissimilar steel joints were analyzed using the two-way thermal coupling finite element method to establish initial stress boundary conditions for simulating SCC propagation influenced by residual stresses. Next, the extended finite element method combined with the maximum principal stress criterion was employed to investigate the propagation direction and path of cracks originating from various initial positions due to welding residual stresses. Concurrently, the J-integral method was used to calculate the stress intensity factors of cracks at different depths along the propagation path. Finally, based on the modified Shoji model, the relationship between the stress intensity factor and SCC propagation rate was examined, allowing for predictions of crack propagation rates and service life for SCC with varying initial defects.
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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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