Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel

Demkowicz, Paul A.; Hunn, John D.; Petti, David A.; Morris, Robert N.

doi:10.1016/j.nucengdes.2017.09.005

Title: Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel

Abstract

The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3 × 105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time-average, volume-average irradiation temperatures of the individual compacts ranged from 955 to 1136 °C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior. The fuel exhibited zero TRISO coating failures (failure of all three dense coating layers) during irradiation and post-irradiation safety testing at temperatures up to 1700 °C. Advanced PIE methods have allowed particles with SiC coating failure that were discovered to be present in a very-low population to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level ofmore »« less

Authors:

Demkowicz, Paul A. ^[1]; Hunn, John D. ^[2]; Petti, David A. ^[1]; Morris, Robert N. ^[2]

Idaho National Lab. (INL), Idaho Falls, ID (United States)
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Publication Date:: Sun Sep 10 00:00:00 EDT 2017

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

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

OSTI Identifier:: 1435299

Alternate Identifier(s):: OSTI ID: 1776132

Grant/Contract Number:: AC05-00OR22725

Resource Type:: Accepted Manuscript

Journal Name:: Nuclear Engineering and Design

Additional Journal Information:: Journal Volume: 329; Journal Issue: C; Journal ID: ISSN 0029-5493

Publisher:: Elsevier

Country of Publication:: United States

Language:: English

Subject:: 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS

Citation Formats


                    Demkowicz, Paul A., Hunn, John D., Petti, David A., and Morris, Robert N. Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel.  United States: N. p., 2017. 
Web.  doi:10.1016/j.nucengdes.2017.09.005.

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                    Demkowicz, Paul A., Hunn, John D., Petti, David A., & Morris, Robert N. Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel.  United States.  https://doi.org/10.1016/j.nucengdes.2017.09.005

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                    Demkowicz, Paul A., Hunn, John D., Petti, David A., and Morris, Robert N. Sun .  
"Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel".  United States.  https://doi.org/10.1016/j.nucengdes.2017.09.005.  https://www.osti.gov/servlets/purl/1435299.

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@article{osti_1435299,

  title        = {Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel},

  author       = {Demkowicz, Paul A. and Hunn, John D. and Petti, David A. and Morris, Robert N.},

  abstractNote = {The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3 × 105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time-average, volume-average irradiation temperatures of the individual compacts ranged from 955 to 1136 °C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior. The fuel exhibited zero TRISO coating failures (failure of all three dense coating layers) during irradiation and post-irradiation safety testing at temperatures up to 1700 °C. Advanced PIE methods have allowed particles with SiC coating failure that were discovered to be present in a very-low population to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of lessons learned from AGR-1 to fuel fabrication and post-irradiation examination for subsequent fuel irradiation experiments as part of the US fuel program are also discussed.},

  doi          = {10.1016/j.nucengdes.2017.09.005},

  journal      = {Nuclear Engineering and Design},

  number       = C,

  volume       = 329,

  place        = {United States},

  year         = {Sun Sep 10 00:00:00 EDT 2017},

  month        = {Sun Sep 10 00:00:00 EDT 2017}

}

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https://doi.org/10.1016/j.nucengdes.2017.09.005

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Works referenced in this record:

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Influence of SiC grain boundary character on fission product transport in irradiated TRISO fuel
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Journal of Nuclear Materials, Vol. 473
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Works referencing / citing this record:

A novel method to inspect coating thickness of tristructural isotropic fuel particles
journal, February 2019

Guo, Man-shan; Yang, Xu; Zhang, Feng
International Journal of Energy Research, Vol. 43, Issue 6
DOI: 10.1002/er.4392

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Similar Records

Title: Key results from irradiation and post-irradiation examination of AGR-1 UCO TRISO fuel

Abstract

Citation Formats

Diffusion of Silver and Cesium in Silicon-Carbide Coatings of Fuel Particles for High-Temperature Gas-Cooled Reactors journal, June 1983

Comparison of silver, cesium, and strontium release predictions using PARFUME with results from the AGR-1 irradiation experiment journal, November 2015

Irradiation performance of AGR-1 high temperature reactor fuel journal, September 2016

First high temperature safety tests of AGR-1 TRISO fuel with the Fuel Accident Condition Simulator (FACS) furnace journal, September 2015

Accident testing of high-temperature reactor fuel elements from the HFR-EU1bis irradiation journal, August 2011

SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products journal, November 2016

Detection and analysis of particles with failed SiC in AGR-1 fuel compacts journal, September 2016

Increase in pyrolytic carbon optical anisotropy and density during processing of coated particle fuel due to heat treatment journal, March 2008

Associations of Pd, U and Ag in the SiC layer of neutron-irradiated TRISO fuel journal, May 2015

Influence of SiC grain boundary character on fission product transport in irradiated TRISO fuel journal, May 2016

Advantages and disadvantages of using a focused ion beam to prepare TEM samples from irradiated U–10Mo monolithic nuclear fuel journal, May 2012

Fission product behavior in Triso-coated UO2 fuel particles journal, February 1994

Fission product palladium-silicon carbide interaction in htgr fuel particles journal, July 1990

Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests journal, September 2016

Silver Release from Coated Particle Fuel journal, September 1977

A novel approach to fabricating fuel compacts for the next generation nuclear plant (NGNP) journal, October 2008

The DOE advanced gas reactor fuel development and qualification program journal, September 2010

Microscopic analysis of irradiated AGR-1 coated particle fuel compacts journal, May 2014

Electron microscopic evaluation and fission product identification of irradiated TRISO coated particles from the AGR-1 experiment: A preliminary review journal, May 2014

Identification of silver and palladium in irradiated TRISO coated particles of the AGR-1 experiment journal, March 2014

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Diffusion of Silver and Cesium in Silicon-Carbide Coatings of Fuel Particles for High-Temperature Gas-Cooled Reactors
journal, June 1983

Comparison of silver, cesium, and strontium release predictions using PARFUME with results from the AGR-1 irradiation experiment
journal, November 2015

Irradiation performance of AGR-1 high temperature reactor fuel
journal, September 2016

First high temperature safety tests of AGR-1 TRISO fuel with the Fuel Accident Condition Simulator (FACS) furnace
journal, September 2015

Accident testing of high-temperature reactor fuel elements from the HFR-EU1bis irradiation
journal, August 2011

SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products
journal, November 2016

Detection and analysis of particles with failed SiC in AGR-1 fuel compacts
journal, September 2016

Increase in pyrolytic carbon optical anisotropy and density during processing of coated particle fuel due to heat treatment
journal, March 2008

Associations of Pd, U and Ag in the SiC layer of neutron-irradiated TRISO fuel
journal, May 2015

Influence of SiC grain boundary character on fission product transport in irradiated TRISO fuel
journal, May 2016

Advantages and disadvantages of using a focused ion beam to prepare TEM samples from irradiated U–10Mo monolithic nuclear fuel
journal, May 2012

Fission product behavior in Triso-coated UO2 fuel particles
journal, February 1994

Fission product palladium-silicon carbide interaction in htgr fuel particles
journal, July 1990

Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests
journal, September 2016

Silver Release from Coated Particle Fuel
journal, September 1977

A novel approach to fabricating fuel compacts for the next generation nuclear plant (NGNP)
journal, October 2008

The DOE advanced gas reactor fuel development and qualification program
journal, September 2010

Microscopic analysis of irradiated AGR-1 coated particle fuel compacts
journal, May 2014

Electron microscopic evaluation and fission product identification of irradiated TRISO coated particles from the AGR-1 experiment: A preliminary review
journal, May 2014

Identification of silver and palladium in irradiated TRISO coated particles of the AGR-1 experiment
journal, March 2014

Coated Particle fuel for high Temperature gas Cooled Reactors
journal, October 2007

A novel method to inspect coating thickness of tristructural isotropic fuel particles
journal, February 2019