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Title: The AGR-3/4 fission product transport irradiation experiment

Abstract

AGR-3/4 was the combined third and fourth planned irradiations for the U.S. Department of Energy Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The primary purpose of the AGR program is to support the development and qualification of tristructural isotropic (TRISO) coated particle fuel for use in High Temperature Gas-cooled Reactors. AGR-3/4 was designed as a fission product transport irradiation experiment whose specific objectives were to: (1) irradiate fuel containing UCO (uranium oxycarbide) designed-to-fail (DTF) fuel particles that provide a fixed source of fission products for subsequent transport through compact matrix and structural graphite materials, (2) assess the effects of sweep gas impurities on fuel performance and fission product transport, (3) provide irradiated fuel and material samples for post-irradiation examination and post-irradiation heating, and (4) support the refinement of fuel performance and fission product transport models. The AGR-3/4 test train was irradiated in the northeast flux trap of the Advanced Test Reactor at Idaho National Laboratory for 369.1 effective full power days from December 2011 to April 2014. The experiment was successful in achieving its specification goals in terms of burnup and fast fluence levels reached at the end of irradiation and fuel temperature levels maintained throughout irradiation: peakmore » compact burnup reached 15.27% fissions per initial heavy-metal atom and peak compact fast fluence reached 5.32 × 10 25 n/m 2 (E > 0.18 MeV), while the time-average volume-average temperatures of the compacts ranged from 854 to 1345°C. Fission product release to birth ratios reached values in the 10 -4 - 10 -3 range early during irradiation as the DTF particles started to fail. In conclusion, subsequent post-irradiation examination will provide information on fission product distributions in matrix and core graphite materials, enabling refinement of fission product transport models.« less

Authors:
 [1];  [1];  [1]; ORCiD logo [1]; ORCiD logo [1]; ORCiD logo [1]; ORCiD logo [1]; ORCiD logo [1]
  1. Idaho National Lab. (INL), Idaho Falls, ID (United States)
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1477031
Alternate Identifier(s):
OSTI ID: 1548863
Report Number(s):
INL/JOU-17-42937-Rev000
Journal ID: ISSN 0029-5493
Grant/Contract Number:  
AC07-05ID14517
Resource Type:
Journal Article: Accepted Manuscript
Journal Name:
Nuclear Engineering and Design
Additional Journal Information:
Journal Volume: 327; 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; post-irradiation examination; TRISO; High Temperature Gas-cooled Reactors; Advanced Gas Reactor; designed-to-fail

Citation Formats

Collin, Blaise P., Demkowicz, Paul A., Petti, David A., Hawkes, Grant L., Palmer, Joe, Pham, Binh T., Scates, Dawn M., and Sterbentz, James W. The AGR-3/4 fission product transport irradiation experiment. United States: N. p., 2017. Web. doi:10.1016/j.nucengdes.2017.12.016.
Collin, Blaise P., Demkowicz, Paul A., Petti, David A., Hawkes, Grant L., Palmer, Joe, Pham, Binh T., Scates, Dawn M., & Sterbentz, James W. The AGR-3/4 fission product transport irradiation experiment. United States. doi:10.1016/j.nucengdes.2017.12.016.
Collin, Blaise P., Demkowicz, Paul A., Petti, David A., Hawkes, Grant L., Palmer, Joe, Pham, Binh T., Scates, Dawn M., and Sterbentz, James W. Sat . "The AGR-3/4 fission product transport irradiation experiment". United States. doi:10.1016/j.nucengdes.2017.12.016. https://www.osti.gov/servlets/purl/1477031.
@article{osti_1477031,
title = {The AGR-3/4 fission product transport irradiation experiment},
author = {Collin, Blaise P. and Demkowicz, Paul A. and Petti, David A. and Hawkes, Grant L. and Palmer, Joe and Pham, Binh T. and Scates, Dawn M. and Sterbentz, James W.},
abstractNote = {AGR-3/4 was the combined third and fourth planned irradiations for the U.S. Department of Energy Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The primary purpose of the AGR program is to support the development and qualification of tristructural isotropic (TRISO) coated particle fuel for use in High Temperature Gas-cooled Reactors. AGR-3/4 was designed as a fission product transport irradiation experiment whose specific objectives were to: (1) irradiate fuel containing UCO (uranium oxycarbide) designed-to-fail (DTF) fuel particles that provide a fixed source of fission products for subsequent transport through compact matrix and structural graphite materials, (2) assess the effects of sweep gas impurities on fuel performance and fission product transport, (3) provide irradiated fuel and material samples for post-irradiation examination and post-irradiation heating, and (4) support the refinement of fuel performance and fission product transport models. The AGR-3/4 test train was irradiated in the northeast flux trap of the Advanced Test Reactor at Idaho National Laboratory for 369.1 effective full power days from December 2011 to April 2014. The experiment was successful in achieving its specification goals in terms of burnup and fast fluence levels reached at the end of irradiation and fuel temperature levels maintained throughout irradiation: peak compact burnup reached 15.27% fissions per initial heavy-metal atom and peak compact fast fluence reached 5.32 × 1025 n/m2 (E > 0.18 MeV), while the time-average volume-average temperatures of the compacts ranged from 854 to 1345°C. Fission product release to birth ratios reached values in the 10-4 - 10-3 range early during irradiation as the DTF particles started to fail. In conclusion, subsequent post-irradiation examination will provide information on fission product distributions in matrix and core graphite materials, enabling refinement of fission product transport models.},
doi = {10.1016/j.nucengdes.2017.12.016},
journal = {Nuclear Engineering and Design},
issn = {0029-5493},
number = C,
volume = 327,
place = {United States},
year = {2017},
month = {12}
}

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