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Fretting Wear Behavior of APMT Steel at 350°C for Reactor Fuel Cladding Application

Published online by Cambridge University Press:  29 July 2016

Thomas Winter ,
James Huggins ,
Richard Neu ,
Preet Singh  and
Chaitanya S. Deo
Thomas Winter*
Affiliation:
Nuclear & Radiological Engineering Program, Georgia Institute of Technology, Atlanta, GA, United States
James Huggins
Affiliation:
George W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA, United States
Richard Neu
Affiliation:
George W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA, United States
Preet Singh
Affiliation:
School of Materials Science and Engineering, Georgia Institute of Technology, Atlanta, GA, United States
Chaitanya S. Deo
Affiliation:
Nuclear & Radiological Engineering Program, Georgia Institute of Technology, Atlanta, GA, United States
*
*(Email: wintertc@gatech.edu)
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Abstract

In support of a recent surge in research to develop an accident tolerant reactor, accident tolerant fuels and cladding candidates are being investigated. Relative motion between the fuel rods and fuel assembly spacer grids can lead to excessive fuel rod wear and, in some cases, to fuel rod failure. Based on industry data, grid-to-rod-fretting (GTRF) has been the number one cause of fuel failures within the U.S. pressurized water reactor (PWR) fleet, accounting for more than 70% of all PWR leaking fuel assemblies. APMT, an Fe-Cr-Al steel alloy, is being examined for the I2S-LWR project as a possible alternative to conventional fuel cladding in a nuclear reactor due to its favorable performance under LOCA conditions. Tests were performed to examine the reliability of the cladding candidate under simulated fretting conditions of a pressurized water reactor (PWR). The contact is simulated with a rectangular and a cylindrical specimen over a line contact area. A combination of SEM analysis and wear & work rate calculations are performed on the samples to determine their performance and wear under fretting. While APMT can perform favorably in loss of coolant accident scenarios, it also needs to perform well when compared to Zircaloy-4 with respect to fretting wear.

Keywords

nuclear materialssteelalloy
Type
Articles
Information
MRS Advances , Volume 1 , Issue 35: Materials Design , 2016 , pp. 2495 - 2500
Copyright
Copyright © Materials Research Society 2016 

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References

REFERENCES

Terrani, K. A., et al. (2014). “Advanced oxidation-resistant iron-based alloys for LWR fuel cladding.” Journal of Nuclear Materials 448(1–3): 420–435.CrossRefGoogle Scholar
Edsinger, , Nucl. News 53 (2010) 40.Google Scholar
Fisher, N. J., et al. (2002). “Fretting-wear of zirconium alloys.” Nuclear Engineering and Design 213(1): 79–90.CrossRefGoogle Scholar
Blau, P. J. (2014). “A multi-stage wear model for grid-to-rod fretting of nuclear fuel rods.” Wear 313(1–2): 89–96.CrossRefGoogle Scholar
Cho, K. H., et al. (1998). “Fretting wear characteristics of Zircaloy-4 tube.” Wear 219(1): 3–7.CrossRefGoogle Scholar
Kim, H. K., et al. (2001). “Fretting wear of laterally supported tube.” Wear 250-251(PART 1): 535–543.Google Scholar
Guérout, F. M., Young, M. Y., Fisher, N. J., & King, S. J. (2005). Fretting-Wear Behavior of Zircaloy-4, OPTIN™, and ZIRLO™ Fuel Rods and Grid Supports Under Various Autoclave and Hydraulic Loop Endurance Test Conditions. Journal of ASTM International, 2(8), 1–24.Google Scholar