Deformation mechanism-based true-stress creep model for SA508 Gr.3 steel over the temperature range of 450e750  C Chuanyang Lu a, b , Xijia Wu c , Yanming He a, d, * , Zengliang Gao a, d , Rong Liu b , Ze Chen a , Wenjian Zheng a , Jianguo Yang a, d a Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, 310014, PR China b Department of Mechanical and Aerospace Engineering, Carleton University, Ottawa, ON, Canada, K1S 5B6 c Structures and Materials Performance Laboratory, Institute for Aerospace Research, National Research Council Canada, Ottawa, ON, Canada, K1A 0R6 d Engineering Research Center of Process Equipment and Remanufacturing, Ministry of Education, China article info Article history: Received 10 July 2019 Received in revised form 20 August 2019 Accepted 29 August 2019 Available online 30 August 2019 Keywords: Reactor pressure vessel In-vessel retention SA508 Gr.3 steel Creep deformation mechanism Creep model abstract Stress and temperature effects on the creep behaviors and mechanisms of a typical nuclear reactor pressure vessel material, SA508 Gr.3 steel, are investigated, over the temperature range of 450e750  C and stress range of 10e400 MPa. Because of the importance of creep life prediction for nuclear reactor failure prevention, three creep models are assessed: Orr-Sherby-Dorn (OSD) and Larson-Miller (LM) parameter methods, and deformation-mechanism based true-stress (DMTS) model. The OSD model employs a single activation energy Q and stress exponent n, which shows a large discrepancy between the experimental and predicted time-to-strain (3% and 5%) data with a coefficient of determination (R 2 ) less than 0.33 over the temperature range of 450e750  C. Both the OSD and LM methods are effective in correlating the time-to-rupture with R 2 ~0.84 over a narrow temperature range of 650e750  C. The DMTS creep model, on the other hand, characterizes the creep behavior in three normalized stress regions: low, intermediate and high, as dominated by grain boundary sliding (GBS), intragranular dislocation climb (IDC) and dislocation glide (IDG), respectively. The microstructural characteristics and creep damage mechanisms of SA508 Gr.3 steel are also examined using scanning/transmission electron microscopy to confirm the predominance of the aforementioned creep deformation mechanisms. The DMTS model provides a fully consistent description of the strain-time curves, the minimum creep rates (MCRs) and the time-to-strain/rupture as well. They are all in good agreement with the experimental observations, particularly with R 2 ~0.94, 0.995, and 0.79 for time to 3% strain, time to 5% strain, and time to rupture, respectively. These analyses demonstrate that the DMTS model is an effective tool in assessing creep properties of SA508 Gr.3 steel for in-vessel retention. © 2019 Elsevier B.V. All rights reserved. 1. Introduction Nuclear power has been an economic, sustainable and environmentally-friendly energy resource, as an alternative to fossil fuels, to meet the increased energy demands over the past several decades [1]. The structural integrity of nuclear reactors under extreme conditions was brought into attention after the Fukushima nuclear accident induced by tsunami in 2011. The reactor pressure vessel (RPV) was destroyed due to the melting of the nuclear cores under the remaining pressure, leading to the failure of nuclear safety barriers [2e4]. To prevent the leakage of radioactive materials from melting cores, the structural integrity of RPVs has to be maintained at all times. For this reason, the concept of in-vessel retention (IVR) has been developed and stressed in the Generation-III light-water reactors, which has been proven to be a successful mitigation strategy in severe nuclear accident [5e8]. The idea of IVR is to flood the reactor cavity with external cooling water when nuclear accident happens. The heat of melting core in the RPV interior will be continuously removed through the external cooling of the reactor [9]. Under such conditions, the temperature at the inner wall of a RPV can reach to ~1300  C due to formation of the melting pool, while the outer wall will be maintained at the * Corresponding author. Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, 310014, PR China. E-mail address: heyanming@zjut.edu.cn (Y. He). Contents lists available at ScienceDirect Journal of Nuclear Materials journal homepage: www.elsevier.com/locate/jnucmat https://doi.org/10.1016/j.jnucmat.2019.151776 0022-3115/© 2019 Elsevier B.V. All rights reserved. Journal of Nuclear Materials 526 (2019) 151776