Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests
Abstract
The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if themore »
- Authors:
-
[1];
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Publication Date:
- Research Org.:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- Sponsoring Org.:
- USDOE Office of Nuclear Energy (NE)
- OSTI Identifier:
- 1325505
- Alternate Identifier(s):
- OSTI ID: 1359688
- Grant/Contract Number:
- AC05-00OR22725
- Resource Type:
- Accepted Manuscript
- Journal Name:
- Nuclear Engineering and Design
- Additional Journal Information:
- Journal Volume: 306; Journal Issue: C; Journal ID: ISSN 0029-5493
- Publisher:
- Elsevier
- Country of Publication:
- United States
- Language:
- English
- Subject:
- 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; TRISO; coated particle fuel; AGR; PIE; tri-structural isotropic (TRISO); fuel; accident testing; fission product release; high-temperature gas-cooled reactor (HTGR); uranium oxide/uranium carbide (UCO) fuel
Citation Formats
Morris, Robert N., Baldwin, Charles A., Demkowicz, Paul A., Hunn, John D., and Reber, Edward L. Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests. United States: N. p., 2016.
Web. doi:10.1016/j.nucengdes.2016.04.031.
Morris, Robert N., Baldwin, Charles A., Demkowicz, Paul A., Hunn, John D., & Reber, Edward L. Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests. United States. https://doi.org/10.1016/j.nucengdes.2016.04.031
Morris, Robert N., Baldwin, Charles A., Demkowicz, Paul A., Hunn, John D., and Reber, Edward L. Wed .
"Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests". United States. https://doi.org/10.1016/j.nucengdes.2016.04.031. https://www.osti.gov/servlets/purl/1325505.
@article{osti_1325505,
title = {Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests},
author = {Morris, Robert N. and Baldwin, Charles A. and Demkowicz, Paul A. and Hunn, John D. and Reber, Edward L.},
abstractNote = {The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.},
doi = {10.1016/j.nucengdes.2016.04.031},
journal = {Nuclear Engineering and Design},
number = C,
volume = 306,
place = {United States},
year = {Wed May 18 00:00:00 EDT 2016},
month = {Wed May 18 00:00:00 EDT 2016}
}
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