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51316-7452-Passivation Characteristics of Ferritic Stainless Materials in Simulated Reactor Environments

Picture for Passivation Characteristics of Ferritic Stainless Materials in Simulated Reactor Environments
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Product Number: 51316-7452-SG
ISBN: 7452 2016 CP
Author: Raul Rebak
Publication Date: 2016
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The US Department of Energy is funding research related to improving the performance of nuclear fuel both under normal operation conditions and beyond design (accident) conditions. A wide variety of cladding candidate materials were tested in high temperature water forone year under simulated normal operation conditions to determine their passivation and corrosion characteristics both for boiling water reactor and pressurized water reactor environments. Materials included Zircaloy-2 T91 nanoferritic and APMT. Results show that the candidate ferritic alloys have in general a better corrosion resistance in high temperature water that the current zirconium based alloy. The oxide film developed on the surface is also thinner in the ferritic materials than in the zirconium alloy.
The US Department of Energy is funding research related to improving the performance of nuclear fuel both under normal operation conditions and beyond design (accident) conditions. A wide variety of cladding candidate materials were tested in high temperature water forone year under simulated normal operation conditions to determine their passivation and corrosion characteristics both for boiling water reactor and pressurized water reactor environments. Materials included Zircaloy-2 T91 nanoferritic and APMT. Results show that the candidate ferritic alloys have in general a better corrosion resistance in high temperature water that the current zirconium based alloy. The oxide film developed on the surface is also thinner in the ferritic materials than in the zirconium alloy.
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