Semi-integral LOCA test of cold-spray chromium coated zircaloy-4 accident tolerant fuel cladding

Hazan, J.; Gauthier, A.; Pouillier, E.; Shirvan, K.

doi:10.1016/j.jnucmat.2021.152940

Title: Semi-integral LOCA test of cold-spray chromium coated zircaloy-4 accident tolerant fuel cladding

Abstract

Enhancing the accident tolerance of PWRs came under serious consideration following the 2011 Fukushima Daiichi nuclear power plant events. While the current nuclear fuels are based on mature technology and have an excellent operational record, further enhancing their strength through accident tolerant fuel (ATF) research and development continues to be of interest. Particularly, the chromium coated clad ATF technology represents an evolutionary change to improve the high temperature performance of existing Zircaloy claddings. Additionally, despite several years of investigation of coated cladding, the integral loss of coolant accident (LOCA) performance of chromium coated cladding has had little attention. In this work, cold-spray chromium coated zircaloy-4 claddings are evaluated using EDF's semi-integral LOCA test facility. Results show that the chromium coating induces a reduced ballooning and burst and an increased burst temperature in the investigated conditions. However, these tests exhibit limited benefits concerning the post-quench behavior of the coated specimens which occurs in the last stage of a prototypical LOCA scenario.

Authors:

Hazan, J. ^[1]; Gauthier, A. ^[2]; Pouillier, E. ^[2];

^[3]

EDF, R&D, Département MMC, Moret-sur-Loing (France); EDF, DPNT, DCN, Saint-Denis (France)
EDF, R&D, Département MMC, Moret-sur-Loing (France)
Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Dept. of Nuclear Science and Engineering

Publication Date:: Thu Mar 11 00:00:00 EST 2021

Research Org.:: Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

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

OSTI Identifier:: 1850475

Alternate Identifier(s):: OSTI ID: 1782290

Grant/Contract Number:: NE0008416

Resource Type:: Accepted Manuscript

Journal Name:: Journal of Nuclear Materials

Additional Journal Information:: Journal Volume: 550; Journal Issue: C; Journal ID: ISSN 0022-3115

Publisher:: Elsevier

Country of Publication:: United States

Language:: English

Subject:: 36 MATERIALS SCIENCE; 73 NUCLEAR PHYSICS AND RADIATION PHYSICS; Materials Science; Nuclear Science & Technology

Citation Formats


                    Hazan, J., Gauthier, A., Pouillier, E., and Shirvan, K. Semi-integral LOCA test of cold-spray chromium coated zircaloy-4 accident tolerant fuel cladding.  United States: N. p., 2021. 
Web.  doi:10.1016/j.jnucmat.2021.152940.

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                    Hazan, J., Gauthier, A., Pouillier, E., & Shirvan, K. Semi-integral LOCA test of cold-spray chromium coated zircaloy-4 accident tolerant fuel cladding.  United States.  https://doi.org/10.1016/j.jnucmat.2021.152940

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                    Hazan, J., Gauthier, A., Pouillier, E., and Shirvan, K. Thu .  
"Semi-integral LOCA test of cold-spray chromium coated zircaloy-4 accident tolerant fuel cladding".  United States.  https://doi.org/10.1016/j.jnucmat.2021.152940.  https://www.osti.gov/servlets/purl/1850475.

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@article{osti_1850475,

  title        = {Semi-integral LOCA test of cold-spray chromium coated zircaloy-4 accident tolerant fuel cladding},

  author       = {Hazan, J. and Gauthier, A. and Pouillier, E. and Shirvan, K.},

  abstractNote = {Enhancing the accident tolerance of PWRs came under serious consideration following the 2011 Fukushima Daiichi nuclear power plant events. While the current nuclear fuels are based on mature technology and have an excellent operational record, further enhancing their strength through accident tolerant fuel (ATF) research and development continues to be of interest. Particularly, the chromium coated clad ATF technology represents an evolutionary change to improve the high temperature performance of existing Zircaloy claddings. Additionally, despite several years of investigation of coated cladding, the integral loss of coolant accident (LOCA) performance of chromium coated cladding has had little attention. In this work, cold-spray chromium coated zircaloy-4 claddings are evaluated using EDF's semi-integral LOCA test facility. Results show that the chromium coating induces a reduced ballooning and burst and an increased burst temperature in the investigated conditions. However, these tests exhibit limited benefits concerning the post-quench behavior of the coated specimens which occurs in the last stage of a prototypical LOCA scenario.},

  doi          = {10.1016/j.jnucmat.2021.152940},

  journal      = {Journal of Nuclear Materials},

  number       = C,

  volume       = 550,

  place        = {United States},

  year         = {Thu Mar 11 00:00:00 EST 2021},

  month        = {Thu Mar 11 00:00:00 EST 2021}

}

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Works referenced in this record:

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Nagase, Fumihisa; Fuketa, Toyoshi
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Fracture Behavior of Irradiated Zircaloy-4 Cladding under Simulated LOCA Conditions
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Nagase, Fumihisa; Fuketa, Toyoshi
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Evaluation of Equivalent Cladding Reacted parameters of Cr-coated claddings oxidized in steam at 1200 °C in relation with oxygen diffusion/partitioning and post-quench ductility
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Brachet, J. C.; Le Saux, M.; Bischoff, J.
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Thieurmel, Ronan; Besson, Jacques; Pouillier, Edouard
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Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors
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Brachet, Jean-Christophe; Idarraga-Trujillo, Isabel; Flem, Marion Le
Journal of Nuclear Materials, Vol. 517
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Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions
journal, December 2016

Park, Dong Jun; Kim, Hyun Gil; Jung, Yang Il
Journal of Nuclear Materials, Vol. 482
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Similar Records

Title: Semi-integral LOCA test of cold-spray chromium coated zircaloy-4 accident tolerant fuel cladding

Abstract

Citation Formats

Effect of Pre-Hydriding on Thermal Shock Resistance of Zircaloy-4 Cladding under Simulated Loss-of-Coolant Accident Conditions journal, July 2004

Behavior of Pre-hydrided Zircaloy-4 Cladding under Simulated LOCA Conditions journal, February 2005

High temperature oxidation and microstructural evolution of cold spray chromium coatings on Zircaloy-4 in steam environments journal, December 2019

Fracture Behavior of Irradiated Zircaloy-4 Cladding under Simulated LOCA Conditions journal, September 2006

Evaluation of Equivalent Cladding Reacted parameters of Cr-coated claddings oxidized in steam at 1200 °C in relation with oxygen diffusion/partitioning and post-quench ductility journal, May 2020

Contribution to the understanding of brittle fracture conditions of zirconium alloy fuel cladding tubes during LOCA transient journal, December 2019

Hydrogen permeation in FeCrAl alloys for LWR cladding application journal, June 2015

Fuel performance of chromium-coated zirconium alloy and silicon carbide accident tolerant fuel claddings journal, October 2018

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

Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions journal, December 2016

Effect of Pre-Hydriding on Thermal Shock Resistance of Zircaloy-4 Cladding under Simulated Loss-of-Coolant Accident Conditions
journal, July 2004

Behavior of Pre-hydrided Zircaloy-4 Cladding under Simulated LOCA Conditions
journal, February 2005

High temperature oxidation and microstructural evolution of cold spray chromium coatings on Zircaloy-4 in steam environments
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Fracture Behavior of Irradiated Zircaloy-4 Cladding under Simulated LOCA Conditions
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Evaluation of Equivalent Cladding Reacted parameters of Cr-coated claddings oxidized in steam at 1200 °C in relation with oxygen diffusion/partitioning and post-quench ductility
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