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Title: Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment

Abstract

Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts taken from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The fuel compacts examined in this study contained TRISO-coated particles with either uranium oxide (UO2) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.1% FIMA. These examinations are intended to explore kernel and coating morphology evolution during irradiation. This includes kernel porosity, swelling, and migration, and irradiation-induced coating fracture and separation. Variations in behavior within a specific cross section, which could be related to temperature or burnup gradients within the fuel compact, are also explored. The criteria for categorizing post-irradiation particle morphologies developed for AGR-1 ceramographic exams, was applied to the particles in the AGR-2 compacts particles examined. Results are compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance.

Authors:
 [1];  [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:
1364473
Report Number(s):
INL/EXT-16-39462
TRN: US1703369
DOE Contract Number:
AC07-05ID14517
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; URANIUM DIOXIDE; URANIUM; TEST REACTORS; URANIUM CARBIDES; CERAMOGRAPHY; COMPARATIVE EVALUATIONS; CROSS SECTIONS; FUEL PARTICLES; AGR; Fuel; TRISO

Citation Formats

Rice, Francine Joyce, and Stempien, John Dennis. Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment. United States: N. p., 2016. Web. doi:10.2172/1364473.
Rice, Francine Joyce, & Stempien, John Dennis. Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment. United States. doi:10.2172/1364473.
Rice, Francine Joyce, and Stempien, John Dennis. 2016. "Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment". United States. doi:10.2172/1364473. https://www.osti.gov/servlets/purl/1364473.
@article{osti_1364473,
title = {Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment},
author = {Rice, Francine Joyce and Stempien, John Dennis},
abstractNote = {Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts taken from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The fuel compacts examined in this study contained TRISO-coated particles with either uranium oxide (UO2) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.1% FIMA. These examinations are intended to explore kernel and coating morphology evolution during irradiation. This includes kernel porosity, swelling, and migration, and irradiation-induced coating fracture and separation. Variations in behavior within a specific cross section, which could be related to temperature or burnup gradients within the fuel compact, are also explored. The criteria for categorizing post-irradiation particle morphologies developed for AGR-1 ceramographic exams, was applied to the particles in the AGR-2 compacts particles examined. Results are compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance.},
doi = {10.2172/1364473},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 9
}

Technical Report:

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  • A series of up to seven irradiation experiments are planned for the Advanced Gas Reactor (AGR) Fuel Development and Quantification Program, with irradiation completed at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for the first experiment (i.e., AGR-1) in November 2009 for an effective 620 full power days. The objective of the AGR-1 experiment was primarily to provide lessons learned on the multi-capsule test train design and to provide early data on fuel performance for use in fuel fabrication process development and post-irradiation safety testing data at high temperatures. This report describes the advanced microscopy and micro-analysismore » results on selected AGR-1 coated particles.« less
  • The electron microscopic examination of selected irradiated TRISO coated particles of the AGR-1 experiment of fuel compact 6-3-2 are presented in this report. Compact 6-3-2 refers to the compact in Capsule 6 at level 3 of Stack 2. The fuel used in capsule 6 compacts, are called the “baseline” fuel as it is fabricated with refined coating process conditions used to fabricate historic German fuel, because of its excellent irradiation performance with UO2 kernels. The AGR-1 fuel is however made of low-enriched uranium oxycarbide (UCO). Kernel diameters are approximately 350 µm with a U-235 enrichment of approximately 19.7%. Compact 6-3-2more » has been irradiated to 11.3% FIMA compact average burn-up with a time average, volume average temperature of 1070.2°C and with a compact average fast fluence of 2.38E21 n/cm« less
  • The electron microscopic examination of selected irradiated TRISO coated particles of the AGR-1 experiment of fuel compact 6-3-2 are presented in this report. Compact 6-3-2 refers to the compact in Capsule 6 at level 3 of Stack 2. The fuel used in capsule 6 compacts, are called the “baseline” fuel as it is fabricated with refined coating process conditions used to fabricate historic German fuel, because of its excellent irradiation performance with UO 2 kernels. The AGR-1 fuel is however made of low-enriched uranium oxycarbide (UCO). Kernel diameters are approximately 350 µm with a U-235 enrichment of approximately 19.7%. Compactmore » 6-3-2 has been irradiated to 11.3% FIMA compact average burn-up with a time average, volume average temperature of 1070.2°C and with a compact average fast fluence of 2.38E21 n/cm« less
  • The HT-12 through HT-15 and HT-17 through HT-19 capsules were uninstrumented experiments conducted in the target position of the High Flux Isotope Reactor at Oak Ridge National Laboratory. The experiments were utilized as evaluation tests of fertile particle coating designs and PyC properties and they provide a basis for normalization of fuel particle design and performance models. A total of thirteen different BISO ThO/sub 2/ particle designs were irradiated to fast neutron fluences ranging from 1.8 to 14.6 x 10/sup 25/ n/m/sup 2/ (E > 29 fJ)/sub HTGR/ and heavy metal burnups ranging from 1.0 to 15.8% FIMA at temperaturesmore » of 1025/sup 0/ to 1580/sup 0/C. The results indicated that BISO ThO/sub 2/ particles can be fabricated to survive the most severe large HTGR design irradiation conditions and that anisotropy of the OPyC coating, which is a strong function of coating rate, is one of the most important variables affecting the irradiation performance of BISO coated fertile particles.« less