Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding

Besmann, T. M.; Yamamoto, Y.; Unocic, K. A.

doi:10.13182/NT15-132

Title: Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding

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Abstract

We assessed the thermochemical compatibility of potential replacement cladding materials for zirconium alloys in light water reactors. Considered were FeCrAl steel (similar to Kanthal APMT), Nb-1%Zr (similar to PWC-11), and a hybrid SiC-composite with a metallic barrier layer. The niobium alloy was also seen as requiring an oxidation protective layer, and a diffusion silicide was investigated. Metallic barrier layers for the SiC-composite reviewed included a FeCrAl alloy, Nb-1%Zr, and chromium. Thermochemical calculations were performed to determine oxidation behavior of the materials in steam, and for hybrid SiC-composites possible interactions between the metallic layer and SiC. Additionally, experimental exposures of SiC-alloy reaction couples at 673K, 1073K, and 1273K for 168 h in an inert atmosphere were made and microanalysis performed. Whereas all materials were determined to oxidize under higher oxygen partial pressures in the steam environment, these varied by material with expected protective oxides forming. Finally, the computed and experimental results indicate the formation of liquid phase eutectic in the FeCrAl-SiC system at the higher temperatures.

Authors:

Besmann, T. M. ^[1]; Yamamoto, Y. ^[1]; Unocic, K. A. ^[1]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Publication Date:: Tue Jun 21 00:00:00 EDT 2016

Research Org.:: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Sponsoring Org.:: USDOE Office of Nuclear Energy (NE)

OSTI Identifier:: 1326468

Grant/Contract Number:: AC05-00OR22725

Resource Type:: Accepted Manuscript

Journal Name:: Nuclear Technology

Additional Journal Information:: Journal Volume: 195; Journal Issue: 2; Journal ID: ISSN 0029-5450

Publisher:: American Nuclear Society (ANS)

Country of Publication:: United States

Language:: English

Subject:: 36 MATERIALS SCIENCE; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS

Citation Formats


                    Besmann, T. M., Yamamoto, Y., and Unocic, K. A. Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding.  United States: N. p., 2016. 
Web.  doi:10.13182/NT15-132.

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                    Besmann, T. M., Yamamoto, Y., & Unocic, K. A. Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding.  United States.  https://doi.org/10.13182/NT15-132

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                    Besmann, T. M., Yamamoto, Y., and Unocic, K. A. Tue .  
"Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding".  United States.  https://doi.org/10.13182/NT15-132.  https://www.osti.gov/servlets/purl/1326468.

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@article{osti_1326468,

  title        = {Thermochemical Compatibility and Oxidation Resistance of Advanced LWR Fuel Cladding},

  author       = {Besmann, T. M. and Yamamoto, Y. and Unocic, K. A.},

  abstractNote = {We assessed the thermochemical compatibility of potential replacement cladding materials for zirconium alloys in light water reactors. Considered were FeCrAl steel (similar to Kanthal APMT), Nb-1%Zr (similar to PWC-11), and a hybrid SiC-composite with a metallic barrier layer. The niobium alloy was also seen as requiring an oxidation protective layer, and a diffusion silicide was investigated. Metallic barrier layers for the SiC-composite reviewed included a FeCrAl alloy, Nb-1%Zr, and chromium. Thermochemical calculations were performed to determine oxidation behavior of the materials in steam, and for hybrid SiC-composites possible interactions between the metallic layer and SiC. Additionally, experimental exposures of SiC-alloy reaction couples at 673K, 1073K, and 1273K for 168 h in an inert atmosphere were made and microanalysis performed. Whereas all materials were determined to oxidize under higher oxygen partial pressures in the steam environment, these varied by material with expected protective oxides forming. Finally, the computed and experimental results indicate the formation of liquid phase eutectic in the FeCrAl-SiC system at the higher temperatures.},

  doi          = {10.13182/NT15-132},

  journal      = {Nuclear Technology},

  number       = 2,

  volume       = 195,

  place        = {United States},

  year         = {Tue Jun 21 00:00:00 EDT 2016},

  month        = {Tue Jun 21 00:00:00 EDT 2016}

}

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