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Title: Safety Testing of AGR-2 UCO Compacts 5-2-2, 2-2-2, and 5-4-1

Post-irradiation examination (PIE) is being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2). This effort builds upon the understanding acquired throughout the AGR-1 PIE campaign, and is establishing a database for the different AGR-2 fuel designs. The AGR-2 irradiation experiment included TRISO fuel particles coated at BWX Technologies (BWXT) with a 150-mm-diameter engineering-scale coater. Two coating batches were tested in the AGR-2 irradiation experiment. Batch 93085 had 508-μm-diameter uranium dioxide (UO2) kernels. Batch 93073 had 427-μm-diameter UCO kernels, which is a kernel design where some of the uranium oxide is converted to uranium carbide during fabrication to provide a getter for oxygen liberated during fission and limit CO production. Fabrication and property data for the AGR-2 coating batches have been compiled and compared to those for AGR-1. The AGR-2 TRISO coatings were most like the AGR-1 Variant 3 TRISO deposited in the 50-mm-diameter ORNL lab-scale coater. In both cases argon-dilution of the hydrogen and methyltrichlorosilane coating gas mixture employed to deposit the SiC was used to produce a finer-grain, more equiaxed SiC microstructure. In addition to the fact that AGR-1 fuel had smaller, 350-μm-diameter UCO kernels,more » notable differences in the TRISO particle properties included the pyrocarbon anisotropy, which was slightly higher in the particles coated in the engineering-scale coater, and the exposed kernel defect fraction, which was higher for AGR-2 fuel due to the detected presence of particles with impact damage introduced during TRISO particle handling.« less
Authors:
 [1] ;  [1] ;  [1] ;  [1]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fusion and Materials for Nuclear Systems Division
Publication Date:
OSTI Identifier:
1328314
Report Number(s):
ORNL/TM-2016/423
DOE Contract Number:
AC05-00OR22725
Resource Type:
Technical Report
Research Org:
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
Sponsoring Org:
USDOE Office of Nuclear Energy (NE), Nuclear Reactor Technologies (NE-7). Advanced Reactor Technologies
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; URANIUM DIOXIDE; PYROLYTIC CARBON; SILICON CARBIDES; CARBON MONOXIDE; MINIMIZATION; FABRICATION; ARGON; IRRADIATION; URANIUM CARBIDES; COMPACTS; HYDROGEN; OXYGEN; COMPARATIVE EVALUATIONS; COATED FUEL PARTICLES; DEPOSITION; DESIGN; MIXTURES; SAFETY; TESTING; POST-IRRADIATION EXAMINATION; ANISOTROPY; COATINGS; DAMAGE; DEFECTS; DILUTION; GETTERS; MICROSTRUCTURE; ACCIDENT-TOLERANT NUCLEAR FUELS; AGR TYPE REACTORS; SILANES TRISO; Coated Particle Fuel; AGR; PIE