Comparison of the radial effects of burnup on fast reactor MOX fuel microstructure and solid fission products

Parrish, Riley J.; Cappia, Fabiola; Aitkaliyeva, Assel

doi:10.1016/j.jnucmat.2020.152003

Title: Comparison of the radial effects of burnup on fast reactor MOX fuel microstructure and solid fission products

Authors:: Parrish, Riley J.; Cappia, Fabiola; Aitkaliyeva, Assel

Publication Date:: Wed Apr 01 00:00:00 EDT 2020

Sponsoring Org.:: USDOE

OSTI Identifier:: 1594211

Resource Type:: Publisher's Accepted Manuscript

Journal Name:: Journal of Nuclear Materials

Additional Journal Information:: Journal Name: Journal of Nuclear Materials Journal Volume: 531 Journal Issue: C; Journal ID: ISSN 0022-3115

Publisher:: Elsevier

Country of Publication:: Netherlands

Language:: English

Citation Formats


                    Parrish, Riley J., Cappia, Fabiola, and Aitkaliyeva, Assel. Comparison of the radial effects of burnup on fast reactor MOX fuel microstructure and solid fission products.  Netherlands: N. p., 2020. 
Web.  doi:10.1016/j.jnucmat.2020.152003.

Copy to clipboard


                    Parrish, Riley J., Cappia, Fabiola, & Aitkaliyeva, Assel. Comparison of the radial effects of burnup on fast reactor MOX fuel microstructure and solid fission products.  Netherlands.  https://doi.org/10.1016/j.jnucmat.2020.152003

Copy to clipboard


                    Parrish, Riley J., Cappia, Fabiola, and Aitkaliyeva, Assel. Wed .  
"Comparison of the radial effects of burnup on fast reactor MOX fuel microstructure and solid fission products".  Netherlands.  https://doi.org/10.1016/j.jnucmat.2020.152003.

Copy to clipboard


                    
@article{osti_1594211,

  title        = {Comparison of the radial effects of burnup on fast reactor MOX fuel microstructure and solid fission products},

  author       = {Parrish, Riley J. and Cappia, Fabiola and Aitkaliyeva, Assel},

  abstractNote = {},

  doi          = {10.1016/j.jnucmat.2020.152003},

  journal      = {Journal of Nuclear Materials},

  number       = C,

  volume       = 531,

  place        = {Netherlands},

  year         = {Wed Apr 01 00:00:00 EDT 2020},

  month        = {Wed Apr 01 00:00:00 EDT 2020}

}

Copy to clipboard

Journal Article:

Free Publicly Available Full Text

Accepted Manuscript (Publisher)

Publisher's Version of Record
https://doi.org/10.1016/j.jnucmat.2020.152003

Other availability

Search WorldCat to find libraries that may hold this journal

Citation Metrics:

Cited by: 4 works

Citation information provided by
Web of Science

Save / Share:

Export Metadata

Save to My Library

Works referenced in this record:

Constitution and thermodynamics of the Mo-Ru, Mo-Pd, Ru-Pd and Mo-Ru-Pd systems
journal, September 1989

Kleykamp, H.
Journal of Nuclear Materials, Vol. 167
DOI: 10.1016/0022-3115(89)90424-8

Development and Verifications of Fast Reactor Fuel Design Code CEPTAR
journal, October 2006

Ozawa, T.; Abe, T.
Nuclear Technology, Vol. 156, Issue 1
DOI: 10.13182/NT156-39

Solubility and migration of fission product barium in oxide fuel
journal, January 1979

Sari, C.; Walker, C. T.; Schumacher, G.
Journal of Nuclear Materials, Vol. 79, Issue 1
DOI: 10.1016/0022-3115(79)90455-0

Materials Challenges for Generation IV Nuclear Energy Systems
journal, June 2008

Allen, T. R.; Sridharan, K.; Tan, L.
Nuclear Technology, Vol. 162, Issue 3
DOI: 10.13182/NT08-A3961

Characterization of alloy particles extracted from spent nuclear fuel
journal, January 2012

Cui, D.; Rondinella, V. V.; Fortner, J. A.
Journal of Nuclear Materials, Vol. 420, Issue 1-3
DOI: 10.1016/j.jnucmat.2011.10.015

The chemical state of the fission products in oxide fuels
journal, April 1985

Kleykamp, H.
Journal of Nuclear Materials, Vol. 131, Issue 2-3, p. 221-246
DOI: 10.1016/0022-3115(85)90460-X

The Chemical State of Fission Products in Oxide Fuels at Different Stages of the Nuclear Fuel Cycle
journal, March 1988

Kleykamp, Heiko
Nuclear Technology, Vol. 80, Issue 3
DOI: 10.13182/NT88-A34065

EBSD and TEM characterization of high burn-up mixed oxide fuel
journal, January 2014

Teague, Melissa; Gorman, Brian; Miller, Brandon
Journal of Nuclear Materials, Vol. 444, Issue 1-3
DOI: 10.1016/j.jnucmat.2013.10.037

Microstructural characterization of high burn-up mixed oxide fast reactor fuel
journal, October 2013

Teague, Melissa; Gorman, Brian; King, Jeffrey
Journal of Nuclear Materials, Vol. 441, Issue 1-3
DOI: 10.1016/j.jnucmat.2013.05.067

Behavior of Fission Products Zirconium and Barium in Fast Reactor Fuel Irradiated to High Burnup
journal, September 1999

Sato, Isamu; Furuya, Hirotaka; Arima, Tatsumi
Journal of Nuclear Science and Technology, Vol. 36, Issue 9
DOI: 10.1080/18811248.1999.9726267

Irradiation performance of PFBR MOX fuel after 112GWd/t burn-up
journal, June 2014

Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.
Journal of Nuclear Materials, Vol. 449, Issue 1-3
DOI: 10.1016/j.jnucmat.2014.01.045

Development of a mixed oxide fuel pin performance analysis code “CEDAR”: Models and analyses of fuel pin irradiation behavior
journal, December 2014

Uwaba, Tomoyuki; Mizuno, Tomoyasu; Nemoto, Junichi
Nuclear Engineering and Design, Vol. 280
DOI: 10.1016/j.nucengdes.2014.08.032

A Liquid-Metal Reactor Core Demonstration Experiment Using HT-9
journal, June 1993

Bridges, Alvia E.; Waltar, Alan E.; Leggett, Robert D.
Nuclear Technology, Vol. 102, Issue 3
DOI: 10.13182/NT93-A17034

Fission product behaviour in phenix fuel pins at high burnup
journal, June 1992

Tourasse, M.; Boidron, M.; Pasquet, B.
Journal of Nuclear Materials, Vol. 188
DOI: 10.1016/0022-3115(92)90453-R

Utilization of dual-column focused ion beam and scanning electron microscope for three dimensional characterization of high burn-up mixed oxide fuel
journal, April 2014

Teague, Melissa; Gorman, Brian
Progress in Nuclear Energy, Vol. 72
DOI: 10.1016/j.pnucene.2013.08.006

Characterization of solid fission products in 13.7% FIMA MOX fuel using electron microscopy techniques
journal, October 2019

Parrish, Riley J.; Wright, Karen E.; Winston, Alexander J.
Journal of Nuclear Materials, Vol. 524
DOI: 10.1016/j.jnucmat.2019.06.042

Microstructural modeling of thermal conductivity of high burn-up mixed oxide fuel
journal, January 2014

Teague, Melissa; Tonks, Michael; Novascone, Stephen
Journal of Nuclear Materials, Vol. 444, Issue 1-3
DOI: 10.1016/j.jnucmat.2013.09.035

Radial microstructural evolution in low burnup fast reactor MOX fuel
journal, September 2019

Parrish, Riley J.; Liu, Xiang; Winston, Alexander
Journal of Nuclear Materials, Vol. 523
DOI: 10.1016/j.jnucmat.2019.06.009

Metallic fission-product inclusions in irradiated oxide fuels
journal, February 1968

Bramman, J. I.; Sharpe, R. M.; Thom, D.
Journal of Nuclear Materials, Vol. 25, Issue 2, p. 201-215
DOI: 10.1016/0022-3115(68)90045-7

Fuel performance of hollow pellets for fast breeder reactors
journal, September 2015

Ishimi, Akihiro; Katsuyama, Kozo; Kihara, Yoshiyuki
Journal of Nuclear Science and Technology, Vol. 53, Issue 7
DOI: 10.1080/00223131.2015.1085337

TEM characterization of high burnup fast-reactor MOX fuel
journal, December 2019

Parrish, Riley; Winston, Alexander; Harp, Jason
Journal of Nuclear Materials, Vol. 527
DOI: 10.1016/j.jnucmat.2019.151794

Behavior of metallic fission products in uranium–plutonium mixed oxide fuel
journal, August 1999

Sato, I.; Furuya, H.; Arima, T.
Journal of Nuclear Materials, Vol. 273, Issue 3
DOI: 10.1016/S0022-3115(99)00071-9

Similar Records in DOE PAGES and OSTI.GOV collections:

Microstructural and chemical characterization of medium burnup fast reactor MOX using electron microscopy

Technical Report Cappia, Fabiola ; Harp, Jason

A thorough post-irradiation examination campaign of medium burnup Sodium Fast Reactor (SFR) mixed oxide (MOX) fuel has been undertaken over the past two years to better understand the performance at intermediate burnup of this annular fuel design. Non-destructive examinations as well as destructive examinations of two pins have been recently published by the authors, providing data regarding the performance at macro-scale. The analyses of selected samples have continued to provide more details regarding the overall microstructure and, particularly, the fuel-cladding interaction.
https://doi.org/10.2172/1784764

Full Text Available
Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

Journal Article Uwaba, Tomoyuki ; Ito, Masahiro ; Katsuyama, Kozo ; ... - Journal of Nuclear Materials

The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39 × 1026 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incrementalmore »« less
https://doi.org/10.1016/j.jnucmat.2011.02.052
TEM characterization of high burnup fast-reactor MOX fuel

Journal Article Parrish, Riley ; Winston, Alexander ; Harp, Jason ; ... - Journal of Nuclear Materials
Cited by 8
https://doi.org/10.1016/j.jnucmat.2019.151794
Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

Journal Article Uwaba, Tomoyuki ; Ito, Masahiro ; Mizuno, Tomoyasu ; ... - Journal of Nuclear Materials, 412(3):294-300

The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39E26 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains weremore »« less
https://doi.org/10.1016/j.jnucmat.2011.02.052
Electron microscopy characterization of the fuel-cladding interaction in medium burnup annular fast reactor MOX

Journal Article Cappia, F. ; Winston, A. ; Miller, B. ; ... - Journal of Nuclear Materials

In this work, we present electron microscopy data focused on the fuel-cladding interaction layer in annular fast reactor MOX with HT-9 cladding at medium burnup. In agreement with previous literature data, the volatile fission products Cs, Te and I have migrated radially into the extreme fuel periphery and partially interacted with the cladding. The accumulation of Cs has occurred in the outermost rim of the fuel pellet, where grain recrystallization has also been observed. Significant amounts of Pd have been found in the interaction zone, particularly in the sample taken from the upper half of the fissile column where themore »« less
https://doi.org/10.1016/j.jnucmat.2021.152922

Full Text Available

Similar Records

Title: Comparison of the radial effects of burnup on fast reactor MOX fuel microstructure and solid fission products

Citation Formats

Constitution and thermodynamics of the Mo-Ru, Mo-Pd, Ru-Pd and Mo-Ru-Pd systems journal, September 1989

Development and Verifications of Fast Reactor Fuel Design Code CEPTAR journal, October 2006

Solubility and migration of fission product barium in oxide fuel journal, January 1979

Materials Challenges for Generation IV Nuclear Energy Systems journal, June 2008

Characterization of alloy particles extracted from spent nuclear fuel journal, January 2012

The chemical state of the fission products in oxide fuels journal, April 1985

The Chemical State of Fission Products in Oxide Fuels at Different Stages of the Nuclear Fuel Cycle journal, March 1988

EBSD and TEM characterization of high burn-up mixed oxide fuel journal, January 2014

Microstructural characterization of high burn-up mixed oxide fast reactor fuel journal, October 2013

Behavior of Fission Products Zirconium and Barium in Fast Reactor Fuel Irradiated to High Burnup journal, September 1999

Irradiation performance of PFBR MOX fuel after 112GWd/t burn-up journal, June 2014

Development of a mixed oxide fuel pin performance analysis code “CEDAR”: Models and analyses of fuel pin irradiation behavior journal, December 2014

A Liquid-Metal Reactor Core Demonstration Experiment Using HT-9 journal, June 1993

Fission product behaviour in phenix fuel pins at high burnup journal, June 1992

Utilization of dual-column focused ion beam and scanning electron microscope for three dimensional characterization of high burn-up mixed oxide fuel journal, April 2014

Characterization of solid fission products in 13.7% FIMA MOX fuel using electron microscopy techniques journal, October 2019

Microstructural modeling of thermal conductivity of high burn-up mixed oxide fuel journal, January 2014

Radial microstructural evolution in low burnup fast reactor MOX fuel journal, September 2019

Metallic fission-product inclusions in irradiated oxide fuels journal, February 1968

Fuel performance of hollow pellets for fast breeder reactors journal, September 2015

TEM characterization of high burnup fast-reactor MOX fuel journal, December 2019

Behavior of metallic fission products in uranium–plutonium mixed oxide fuel journal, August 1999

Constitution and thermodynamics of the Mo-Ru, Mo-Pd, Ru-Pd and Mo-Ru-Pd systems
journal, September 1989

Development and Verifications of Fast Reactor Fuel Design Code CEPTAR
journal, October 2006

Solubility and migration of fission product barium in oxide fuel
journal, January 1979

Materials Challenges for Generation IV Nuclear Energy Systems
journal, June 2008

Characterization of alloy particles extracted from spent nuclear fuel
journal, January 2012

The chemical state of the fission products in oxide fuels
journal, April 1985

The Chemical State of Fission Products in Oxide Fuels at Different Stages of the Nuclear Fuel Cycle
journal, March 1988

EBSD and TEM characterization of high burn-up mixed oxide fuel
journal, January 2014

Microstructural characterization of high burn-up mixed oxide fast reactor fuel
journal, October 2013

Behavior of Fission Products Zirconium and Barium in Fast Reactor Fuel Irradiated to High Burnup
journal, September 1999

Irradiation performance of PFBR MOX fuel after 112GWd/t burn-up
journal, June 2014

Development of a mixed oxide fuel pin performance analysis code “CEDAR”: Models and analyses of fuel pin irradiation behavior
journal, December 2014

A Liquid-Metal Reactor Core Demonstration Experiment Using HT-9
journal, June 1993

Fission product behaviour in phenix fuel pins at high burnup
journal, June 1992

Utilization of dual-column focused ion beam and scanning electron microscope for three dimensional characterization of high burn-up mixed oxide fuel
journal, April 2014

Characterization of solid fission products in 13.7% FIMA MOX fuel using electron microscopy techniques
journal, October 2019

Microstructural modeling of thermal conductivity of high burn-up mixed oxide fuel
journal, January 2014

Radial microstructural evolution in low burnup fast reactor MOX fuel
journal, September 2019

Metallic fission-product inclusions in irradiated oxide fuels
journal, February 1968

Fuel performance of hollow pellets for fast breeder reactors
journal, September 2015

TEM characterization of high burnup fast-reactor MOX fuel
journal, December 2019

Behavior of metallic fission products in uranium–plutonium mixed oxide fuel
journal, August 1999