Abstract
Earlier tests (1) carried out in sodium polluted (up to 10%) by sodium hydroxide between 400 and 550 °C have shown that the austenitic materials used (especially Alloy 800) for the steam generator tubes of fast breeder reactors cooled with sodium could induce a risk of stress corrosion cracking in this medium. However, in sodium polluted by sodium oxide, the same materials are not affected by this phenomenon in the same conditions of stress, but are only subject to slight intergranular corrosion.
This is a preview of subscription content, log in via an institution to check access.
Access this chapter
Tax calculation will be finalised at checkout
Purchases are for personal use only
Preview
Unable to display preview. Download preview PDF.
References
P. Berge, C. Oberlin, P. Saint-Paul, M. Zbinden, J. of Nuclear Materials, 66, 65 (1977).
EDF, DER, Internal report (1980).
J. L. Henry, S. A. O’Hare, M. P. Krug, USBM, 1532, (1971).
J. P. Maupré, Report CEA, R 4905 (1978).
Author information
Authors and Affiliations
Editor information
Editors and Affiliations
Rights and permissions
Copyright information
© 1982 Plenum Press, New York
About this chapter
Cite this chapter
Oberlin, C., Saint-Paul, P. (1982). Thermodynamic and Experimental Study of Sodium Hydroxide Decomposition in Sodium between 430 and 550 °C. In: Borgstedt, H.U. (eds) Material Behavior and Physical Chemistry in Liquid Metal Systems. Springer, Boston, MA. https://doi.org/10.1007/978-1-4684-8366-6_28
Download citation
DOI: https://doi.org/10.1007/978-1-4684-8366-6_28
Publisher Name: Springer, Boston, MA
Print ISBN: 978-1-4684-8368-0
Online ISBN: 978-1-4684-8366-6
eBook Packages: Springer Book Archive