US EPA

4724409 

Journal Article 

The AGR-3/4 fission product transport irradiation experiment 

Collin, BP; Demkowicz, PA; Petti, DA; Hawkes, GL; Palmer, Joe; Pham, BT; Scates, DM; Sterbentz, JW 

2018 

1 

Nuclear Engineering and Design
ISSN: 0029-5493
EISSN: 1872-759X 

327 

212-227 

10.1016/j.nucengdes.2017.12.016 

WOS:000425302400019 

AGR-3/4 was the combined third and fourth planned irradiations for the U.S. Department of Energy (DOE) 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 (PIE) 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 (ATR) at Idaho National Laboratory (INL) 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 5.32x10(25) n/m(2) (E > 0.18 MeV), while the time-average volume-average temperatures of the compacts ranged from 854 to 1345 degrees 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. Subsequent post-irradiation examination will provide information on fission product distributions in matrix and core graphite materials, enabling refinement of fission product transport models.