Abstract
The average size and density evolution of dislocation loops in AL-6XN austenitic stainless steel, a candidate fuel cladding material for supercritical water-cooled reactor, under proton irradiation were simulated through a rate theory model. The simulation results exhibit relatively good agreement with the experimental results at 563 K. The size and density of defect clusters are calculated under irradiation temperature between 550 K and 900 K and irradiation doses up to 15 dpa which satisfies the working condition in supercritical water-cooled reactor. The fast nucleation between self-interstitials happens at the initial stage of irradiation. The average size of dislocation loops increases while the average density of these loops reduces with the increasing temperature, and the average density approaches to a constant when irradiated at higher irradiation doses. The mechanism is discussed based on the variation of rate constants of defect reactions and the variation of the diffusion coefficients of interstitials and dislocation loops with dose and temperature.
Publisher
Trans Tech Publications, Ltd.
Subject
Mechanical Engineering,Mechanics of Materials,Condensed Matter Physics,General Materials Science
Reference40 articles.
1. C. Sun, R. Hui, W. Qu, S. Yick, Progress in corrosion resistant materials for supercritical water reactors. Corros. Sci. 51 (2009) 2508-2523.
2. G.S. Was, P. Ampornrat, G. Gupta, S. Teysseyre, E.A. West, T.R. Allen, K. Sridharan, L. Tan, Y. Chen, X. Ren, C. Pister, Corrosion and stress corrosion cracking in supercritical water. J. Nucl. Mater. 371 (2007) 176-201.
3. M. Naidin, S. Mokry, F. Baig, Y. Gospodinov, U. Zirn, I. Pioro, G. Naterer, Thermal-Design Options for Pressure-Channel SCWRS With Cogeneration of Hydrogen. J. Eng. Gas. Turb. Power, 131 (2009) 012901.
4. S. Baindur, Materials challenges for the supercritical water-cooled reactor (SCWR). Bullet. Canad. Nucl. Soci, 29 (2008) 32-38.
5. J. Buongiorno, P. MacDonald, Supercritical water reactor (SCWR). Progress Report for the FY-03 Generation-IV R&D Activities for the Development of the SCWR in the US, INEEL/Ext-03-03-01210, INEEL, USA, September, (2003).
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