Original Article PWSCC Growth Assessment Model Considering Stress Triaxiality Factor for Primary Alloy 600 Components Jong-Sung Kim a,* , Ji-Soo Kim b , Jun-Young Jeon b , and Yun-Jae Kim b a Department of Nuclear Engineering, Sejong University, Gunja-ro, Seoul, KS013, Republic of Korea b Department of Mechanical Engineering, Korea University, Inchon-ro, Seoul, KS013, Republic of Korea article info Article history: Received 14 December 2015 Received in revised form 29 February 2016 Accepted 2 March 2016 Available online 29 March 2016 Keywords: Alloy 600 Primary Water Stress Corrosion Cracking (PWSCC) SCC Growth Simulation Steam Generator Tube Stress Triaxiality abstract We propose a primary water stress corrosion cracking (PWSCC) initiation model of Alloy 600 that considers the stress triaxiality factor to apply to finite element analysis. We investigated the correlation between stress triaxiality effects and PWSCC growth behavior in cold-worked Alloy 600 stream generator tubes, and identified an additional stress triaxiality factor that can be added to Garud's PWSCC initiation model. By applying the proposed PWSCC initiation model considering the stress triaxiality factor, PWSCC growth simulations based on the macroscopic phenomenological damage mechanics approach were carried out on the PWSCC growth tests of various cold-worked Alloy 600 steam generator tubes and compact tension specimens. As a result, PWSCC growth behavior results from the finite element prediction are in good agreement with the experimental results. Copyright © 2016, Published by Elsevier Korea LLC on behalf of Korean Nuclear Society. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/ licenses/by-nc-nd/4.0/). 1. Introduction In pressurized water reactors (PWRs), Alloy 600 has been used as a primary boundary material of penetration nozzles in reactor pressure vessels, steam generator tubes, and similar applications [1]. However, Alloy 600 components in PWR are known to be sensitive to primary water stress corrosion cracking (PWSCC), and the evaluation of PWSCC growth in Alloy 600 components is one of the major issues in assessing the structural integrity of degraded PWRs. In primary Alloy 600 components, the high tensile residual stress and the plastic deformation are generated due to cold work (CW), such as tube expansion or welding processes. It is well known that CW in Alloy 600 generally accelerates the initiation and growth of stress corrosion cracking (SCC) [2e4]. The PWSCC growth assessment procedure for primary Alloy 600 components is presented in the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME B&PV) Code XI [5]. This is based on empirical models such as Scott's * Corresponding author. E-mail address: [email protected] (J.-S. Kim). Available online at ScienceDirect Nuclear Engineering and Technology journal homepage: www.elsevier.com/locate/net Nuclear Engineering and Technology 48 (2016) 1036 e1046 http://dx.doi.org/10.1016/j.net.2016.03.003 1738-5733/Copyright © 2016, Published by Elsevier Korea LLC on behalf of Korean Nuclear Society. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).
PWSCC Growth Assessment Model Considering Stress Triaxiality Factor for Primary Alloy 600 Components
Jun 12, 2023
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