Autoclave grid-to-rod fretting wear evaluation of a candidate cladding coating for accident-tolerant fuel

Reed, Brady; Wang, Rick; Lu, Roger Y.; Qu, Jun

doi:10.1016/j.wear.2020.203578

Title: Autoclave grid-to-rod fretting wear evaluation of a candidate cladding coating for accident-tolerant fuel

Abstract

In pressurized water reactors (PWRs), water flow induced vibrations cause contact and rubbing between the fuel rods and the supporting grid, a phenomenon known as Grid-to-Rod-Fretting (GTRF). GTRF may produce progressive wear damage on the fuel claddings leading to subsequent leakage of radioactive fission products. Various accident-tolerant fuel (ATF) concepts are being developed for higher resistance to the high temperature steam and one approach is to apply a cladding coating. Here, fretting wear behavior of a candidate Cr-coating was investigated using a unique bench-scale autoclave testing rig mimicking the environment in an industrial full-assembly PWR simulator. The contact was under a realistically low load (~0.5 N) lubricated by deionized water at a temperature of 204 °C under a pressure of 20-23 bars. Results demonstrated that the Cr-coating significantly improved the cladding's wear resistance when tested against a commercial ZIRLO grid with or without pre-oxidization. In addition, the Cr-coating also reduced wear on the non-oxidized ZIRLO grid but slightly increased the wear on the pre-oxidized grid.

Authors:

^[1];

^[1]; Lu, Roger Y. ^[2];

^[1]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science & Technology Division
Westinghouse Electric Company, Hopkins, SC (United States)

Publication Date:: Sun Dec 13 00:00:00 EST 2020

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

Sponsoring Org.:: USDOE Office of Nuclear Energy (NE), Nuclear Fuel Cycle and Supply Chain. Advanced Fuel Campaign; USDOE

OSTI Identifier:: 1756262

Alternate Identifier(s):: OSTI ID: 1776259

Grant/Contract Number:: AC05-00OR22725

Resource Type:: Accepted Manuscript

Journal Name:: Wear

Additional Journal Information:: Journal Volume: 466-467; Journal ID: ISSN 0043-1648

Publisher:: Elsevier

Country of Publication:: United States

Language:: English

Subject:: 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; grid-to-rod-fretting (GTRF); accident-tolerant fuel (ATF); Cr-coating; pressurized water reactors (PWRs); autoclave; wear

Citation Formats


                    Reed, Brady, Wang, Rick, Lu, Roger Y., and Qu, Jun. Autoclave grid-to-rod fretting wear evaluation of a candidate cladding coating for accident-tolerant fuel.  United States: N. p., 2020. 
Web.  doi:10.1016/j.wear.2020.203578.

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                    Reed, Brady, Wang, Rick, Lu, Roger Y., & Qu, Jun. Autoclave grid-to-rod fretting wear evaluation of a candidate cladding coating for accident-tolerant fuel.  United States.  https://doi.org/10.1016/j.wear.2020.203578

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                    Reed, Brady, Wang, Rick, Lu, Roger Y., and Qu, Jun. Sun .  
"Autoclave grid-to-rod fretting wear evaluation of a candidate cladding coating for accident-tolerant fuel".  United States.  https://doi.org/10.1016/j.wear.2020.203578.  https://www.osti.gov/servlets/purl/1756262.

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@article{osti_1756262,

  title        = {Autoclave grid-to-rod fretting wear evaluation of a candidate cladding coating for accident-tolerant fuel},

  author       = {Reed, Brady and Wang, Rick and Lu, Roger Y. and Qu, Jun},

  abstractNote = {In pressurized water reactors (PWRs), water flow induced vibrations cause contact and rubbing between the fuel rods and the supporting grid, a phenomenon known as Grid-to-Rod-Fretting (GTRF). GTRF may produce progressive wear damage on the fuel claddings leading to subsequent leakage of radioactive fission products. Various accident-tolerant fuel (ATF) concepts are being developed for higher resistance to the high temperature steam and one approach is to apply a cladding coating. Here, fretting wear behavior of a candidate Cr-coating was investigated using a unique bench-scale autoclave testing rig mimicking the environment in an industrial full-assembly PWR simulator. The contact was under a realistically low load (~0.5 N) lubricated by deionized water at a temperature of 204 °C under a pressure of 20-23 bars. Results demonstrated that the Cr-coating significantly improved the cladding's wear resistance when tested against a commercial ZIRLO grid with or without pre-oxidization. In addition, the Cr-coating also reduced wear on the non-oxidized ZIRLO grid but slightly increased the wear on the pre-oxidized grid.},

  doi          = {10.1016/j.wear.2020.203578},

  journal      = {Wear},

  number       = ,

  volume       = 466-467,

  place        = {United States},

  year         = {Sun Dec 13 00:00:00 EST 2020},

  month        = {Sun Dec 13 00:00:00 EST 2020}

}

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Works referenced in this record:

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation
journal, June 2016

Qu, Jun; Cooley, Kevin M.; Shaw, Austin H.
Wear, Vol. 356-357
DOI: 10.1016/j.wear.2016.02.020

Grid-to-rod flow-induced impact study for PWR fuel in reactor
journal, August 2016

Jiang, Hao; Qu, Jun; Lu, Roger Y.
Progress in Nuclear Energy, Vol. 91
DOI: 10.1016/j.pnucene.2016.06.003

Fretting-wear of zirconium alloys
journal, April 2002

Fisher, Nigel J.; Weckwerth, Murray K.; Grandison, Dwight A. E.
Nuclear Engineering and Design, Vol. 213, Issue 1
DOI: 10.1016/S0029-5493(02)00035-3

Research on performance enhancement of nuclear fuel with SiC cladding by using high thermal conductivity fuels
journal, June 2020

Deng, Yangbin; Qiu, Bowen; Pang, Bo
Progress in Nuclear Energy, Vol. 124
DOI: 10.1016/j.pnucene.2020.103330

A review on thermohydraulic and mechanical-physical properties of SiC, FeCrAl and Ti3SiC2 for ATF cladding
journal, January 2020

Qiu, Bowen; Wang, Jun; Deng, Yangbin
Nuclear Engineering and Technology, Vol. 52, Issue 1
DOI: 10.1016/j.net.2019.07.030

A comparative study on the wear behaviors of cladding candidates for accident-tolerant fuel
journal, October 2015

Lee, Young-Ho; Byun, Thak Sang
Journal of Nuclear Materials, Vol. 465
DOI: 10.1016/j.jnucmat.2015.05.017

Accident tolerant fuel cladding development: Promise, status, and challenges
journal, April 2018

Terrani, Kurt A.
Journal of Nuclear Materials, Vol. 501
DOI: 10.1016/j.jnucmat.2017.12.043

Experimental evaluation of cold spray FeCrAl alloys coated zirconium-alloy for potential accident tolerant fuel cladding
journal, December 2019

Dabney, Tyler; Johnson, Greg; Yeom, Hwasung
Nuclear Materials and Energy, Vol. 21
DOI: 10.1016/j.nme.2019.100715

Effects of amplitude and frequency on the wear mode change of Inconel 690 SG tube mated with SUS 409
journal, May 2014

Yun, Jae Yong; Park, Myung Chul; Shin, Gyeong Su
Wear, Vol. 313, Issue 1-2
DOI: 10.1016/j.wear.2014.02.019

A multi-stage wear model for grid-to-rod fretting of nuclear fuel rods
journal, May 2014

Blau, Peter J.
Wear, Vol. 313, Issue 1-2
DOI: 10.1016/j.wear.2014.02.016

Cracking and spalling of the oxide layer developed in high-burnup Zircaloy-4 cladding under normal operating conditions in a PWR
journal, December 2018

Hong, Kisik; Barber, J. R.; Thouless, M. D.
Journal of Nuclear Materials, Vol. 512
DOI: 10.1016/j.jnucmat.2018.10.003

Effect of a surface oxide-dispersion-strengthened layer on mechanical strength of zircaloy-4 tubes
journal, March 2018

Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan
Nuclear Engineering and Technology, Vol. 50, Issue 2
DOI: 10.1016/j.net.2017.12.001

Enhanced wear resistance of CrAl-coated cladding for accident-tolerant fuel
journal, September 2019

Lee, Young-Ho; Park, Jung-Hwan; Kim, Il-Hyun
Journal of Nuclear Materials, Vol. 523
DOI: 10.1016/j.jnucmat.2019.06.011

AREVA NP's enhanced accident-tolerant fuel developments: Focus on Cr-coated M5 cladding
journal, March 2018

Bischoff, Jeremy; Delafoy, Christine; Vauglin, Christine
Nuclear Engineering and Technology, Vol. 50, Issue 2
DOI: 10.1016/j.net.2017.12.004

Modeling of Complex Wear Behavior Associated with Grid-to-Rod Fretting in Light Water Nuclear Reactors
journal, September 2016

Blau, P. J.; Qu, J.; Lu, R.
JOM, Vol. 68, Issue 11
DOI: 10.1007/s11837-016-2113-9

A study on fretting fatigue characteristics of Inconel 690 at high temperature
journal, October 2011

Kwon, Jae-do; Jeung, Han-kyu; Chung, Il-sup
Tribology International, Vol. 44, Issue 11
DOI: 10.1016/j.triboint.2010.11.006

Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors
journal, April 2019

Brachet, Jean-Christophe; Idarraga-Trujillo, Isabel; Flem, Marion Le
Journal of Nuclear Materials, Vol. 517
DOI: 10.1016/j.jnucmat.2019.02.018

Investigating grid-to-rod fretting wear of nuclear fuel claddings using a unique autoclave fretting rig
journal, October 2018

Lazarevic, Sladjan; Lu, Roger Y.; Favede, Cyrille
Wear, Vol. 412-413
DOI: 10.1016/j.wear.2018.06.011

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Title: Autoclave grid-to-rod fretting wear evaluation of a candidate cladding coating for accident-tolerant fuel

Abstract

Citation Formats

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation journal, June 2016

Grid-to-rod flow-induced impact study for PWR fuel in reactor journal, August 2016

Fretting-wear of zirconium alloys journal, April 2002

Research on performance enhancement of nuclear fuel with SiC cladding by using high thermal conductivity fuels journal, June 2020

A review on thermohydraulic and mechanical-physical properties of SiC, FeCrAl and Ti3SiC2 for ATF cladding journal, January 2020

A comparative study on the wear behaviors of cladding candidates for accident-tolerant fuel journal, October 2015

Accident tolerant fuel cladding development: Promise, status, and challenges journal, April 2018

Experimental evaluation of cold spray FeCrAl alloys coated zirconium-alloy for potential accident tolerant fuel cladding journal, December 2019

Effects of amplitude and frequency on the wear mode change of Inconel 690 SG tube mated with SUS 409 journal, May 2014

A multi-stage wear model for grid-to-rod fretting of nuclear fuel rods journal, May 2014

Cracking and spalling of the oxide layer developed in high-burnup Zircaloy-4 cladding under normal operating conditions in a PWR journal, December 2018

Effect of a surface oxide-dispersion-strengthened layer on mechanical strength of zircaloy-4 tubes journal, March 2018

Enhanced wear resistance of CrAl-coated cladding for accident-tolerant fuel journal, September 2019

AREVA NP's enhanced accident-tolerant fuel developments: Focus on Cr-coated M5 cladding journal, March 2018

Modeling of Complex Wear Behavior Associated with Grid-to-Rod Fretting in Light Water Nuclear Reactors journal, September 2016

A study on fretting fatigue characteristics of Inconel 690 at high temperature journal, October 2011

Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors journal, April 2019

Investigating grid-to-rod fretting wear of nuclear fuel claddings using a unique autoclave fretting rig journal, October 2018

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation
journal, June 2016

Grid-to-rod flow-induced impact study for PWR fuel in reactor
journal, August 2016

Fretting-wear of zirconium alloys
journal, April 2002

Research on performance enhancement of nuclear fuel with SiC cladding by using high thermal conductivity fuels
journal, June 2020

A review on thermohydraulic and mechanical-physical properties of SiC, FeCrAl and Ti3SiC2 for ATF cladding
journal, January 2020

A comparative study on the wear behaviors of cladding candidates for accident-tolerant fuel
journal, October 2015

Accident tolerant fuel cladding development: Promise, status, and challenges
journal, April 2018

Experimental evaluation of cold spray FeCrAl alloys coated zirconium-alloy for potential accident tolerant fuel cladding
journal, December 2019

Effects of amplitude and frequency on the wear mode change of Inconel 690 SG tube mated with SUS 409
journal, May 2014

A multi-stage wear model for grid-to-rod fretting of nuclear fuel rods
journal, May 2014

Cracking and spalling of the oxide layer developed in high-burnup Zircaloy-4 cladding under normal operating conditions in a PWR
journal, December 2018

Effect of a surface oxide-dispersion-strengthened layer on mechanical strength of zircaloy-4 tubes
journal, March 2018

Enhanced wear resistance of CrAl-coated cladding for accident-tolerant fuel
journal, September 2019

AREVA NP's enhanced accident-tolerant fuel developments: Focus on Cr-coated M5 cladding
journal, March 2018

Modeling of Complex Wear Behavior Associated with Grid-to-Rod Fretting in Light Water Nuclear Reactors
journal, September 2016

A study on fretting fatigue characteristics of Inconel 690 at high temperature
journal, October 2011

Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors
journal, April 2019

Investigating grid-to-rod fretting wear of nuclear fuel claddings using a unique autoclave fretting rig
journal, October 2018