Evaluation of Tensile Property of Austenitic Alloys Exposed to High-Temperature S-CO2 Environment

고온 S-CO2 환경에 노출된 오스테나이트계 합금의 인장특성 평가

  • Received : 2014.03.15
  • Accepted : 2014.07.29
  • Published : 2014.12.01


Super-critical $CO_2$ ($S-CO_2$) Brayton cycle has been considered to replace the current steam Rankine cycle in Sodium-cooled Fast Reactor (SFR) in order to improve the inherent safety and thermal efficiency. Several austenitic alloys are considered as the structural materials for high temperature $S-CO_2$ environment.. Microstructural change after long-term exposure to high temperature $S-CO_2$ environment could affect to the mechanical properties. In this study, candidate materials (austenitic stainless steels and Alloy 800HT) were exposed to $S-CO_2$ to assess oxidation resistance and the change in tensile properties. Loss of ductility was observed for some austenitic stainless steels even after 250 h exposure. The contribution of $S-CO_2$ environment on such changes was analyzed based on the characterization of the surface oxide and carburization of the materials in which 316H and 800H showed different oxidation behaviors.


Supercritical Carbon Dioxide;Austenitic Alloys;Corrosion property;Tensile property


  1. Chang, Y. I., Finck, P. J. and Grandy, C., 2006, Advanced Burner Test Reactor Preconceptual Design Report, Argonne National Laboratory report ANL-ABR-1 (ANL-AFCL-173).
  2. Dong, Z., Li, Y., Lin, M. and Li, M., 2013, "A Study of the Mechanism of Enhancing Oil Recovery Using Supercritical Carbon Dioxide Microemulsion," Petroleum Sci., Vol. 30, No. 1, pp. 91-96.
  3. Corradini, M., 2010, Advanced Burner Reactor Sodium Technology Gap Analysis, U.S. DOE Report FCR&D-REAC-2010-000034, Sandia National Laboratories.
  4. Beech, D. J. and May, R., 1999, "Gas Reactor and Associated Nuclear Experience in The UK Relevant to High Temperature Reactor Engineering," Proceedings of the First Information Exchange Meeting on Basic Studies on High-Temperature Engineering, Paris, France.
  5. Nam, H. Y., Kim, J. B., Lee, J. H. and Park, C. G., 2011, "Concept Development and Review of Current Technical Issues for SFR Steam Generator," Trans. Korean Soc. Mech. Eng. A, Vol. 35, pp.1083-1090.
  6. Dostal, V., Driscoll, M. J. and Hejzlar, P., 2004, A Supercritical Carbon Dioxide Cycle for Next Generation Nuclear Reactors, MIT Annual and Progress Reports, MITANP-TR-100.
  7. Pillai, S. R. and Khatak, H. S., 2002, "Corrosion of Austenitic Stainless Steel in Liquid Sodium," Corrosion of Austenitic Stainless Steels: Mechanism, Mitigation and Monitoring, ISBN 1085573-613-6 chapter 10, pp. 241-264.
  8. Natesan, K., Li, M., Chopra, O.K. and Majumdar, S., 2009, "Sodium Effects on Mechanical Performance and Consideration in High Temperature Structural Design for Advanced Reactors," J. of Nucl. Mat., Vol. 392, pp. 243-249.
  9. Cao, G., Firouzdor, V., Sridharan, K., Anderson, M. and Allen, T.R., 2012, "Corrosion of Austenitic Alloys in High Temperature Supercritical Carbon Dioxide," Corrosion Science, Vol. 60, pp. 246-255.
  10. Sridharan, K., 2013, Corrosion in Supercritical Carbon Dioxide: Materials, Environmental Purity, Surface Treatments, and Flow Issues, Final Report 10-872, University of Wisconsin.
  11. Faooq, M., 2013, Strengthening and Degradation Mechanisms in Austenitic Stainless Steels at Elevated Temperature, KTH Sweden Doctoral Thesis.
  12. Was, G., 2013, Corrosion and Creep of Candidate Alloys in High Temperature Helium and Steam Environments for NGNP, U.S. Nuclear Energy University Programs Final report NEUP 09-678, University of Michigan.
  13. Fulger, M., Ohai, D., Mihalache, M., Pantiru, M. and Malinovschi, V., 2009, "Oxidation Behavior of Incoloy 800 under Simulated Supercritical Water Conditions," J. Nucl. Mat. Vol. 385, pp. 288-293.
  14. Moore, R. and Conboy, T., 2012, Metal Corrosion in a Supercritical Carbon Dioxide - Liquid Sodium Power Cycle, Milestone Report, M3AR12SN08010601, Sandia National Laboratories.


Supported by : 미래창조과학부