Pre- and post-irradiation characterization and properties measurements of ZrC coated surrogate TRISO particles

Technical Report · Wed Sep 01 00:00:00 EDT 2010

DOI:https://doi.org/10.2172/989722· OSTI ID:989722

Vasudevamurthy, Gokul ^[1]; Katoh, Yutai ^[1]; Hunn, John D ^[1]; Snead, Lance Lewis ^[1]

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1. ORNL

Zirconium carbide is a candidate to either replace or supplement silicon carbide as a coating material in TRISO fuel particles for high temperature gas-cooled reactor fuels. Six sets of ZrC coated surrogate microsphere samples, fabricated by the Japan Atomic Energy Agency using the fluidized bed chemical vapor deposition method, were irradiated in the High Flux Isotope Reactor at the Oak Ridge National Laboratory. These developmental samples available for the irradiation experiment were in conditions of either as-fabricated coated particles or particles that had been heat-treated to simulate the fuel compacting process. Five sets of samples were composed of nominally stoichiometric compositions, with the sixth being richer in carbon (C/Zr = 1.4). The samples were irradiated at 800 and 1250 C with fast neutron fluences of 2 and 6 dpa. Post-irradiation, the samples were retrieved from the irradiation capsules followed by microstructural examination performed at the Oak Ridge National Laboratory's Low Activation Materials Development and Analysis Laboratory. This work was supported by the US Department of Energy Office of Nuclear Energy's Advanced Gas Reactor program as part of International Nuclear Energy Research Initiative collaboration with Japan. This report includes progress from that INERI collaboration, as well as results of some follow-up examination of the irradiated specimens. Post-irradiation examination items included microstructural characterization, and nanoindentation hardness/modulus measurements. The examinations revealed grain size enhancement and softening as the primary effects of both heat-treatment and irradiation in stoichiometric ZrC with a non-layered, homogeneous grain structure, raising serious concerns on the mechanical suitability of these particular developmental coatings as a replacement for SiC in TRISO fuel. Samples with either free carbon or carbon-rich layers dispersed in the ZrC coatings experienced negligible grain size enhancement during both heat treatment and irradiation. However, these samples experienced irradiation induced softening similar to stoichiometric ZrC samples.

• Research Organization: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: Work for Others (WFO)
• DOE Contract Number: DE-AC05-00OR22725
• OSTI ID: 989722
• Report Number(s): ORNL/TM-2010/223; TRN: US1007235
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
CARBON
CHEMICAL VAPOR DEPOSITION
COATINGS
COMPACTING
FAST NEUTRONS
FLUIDIZED BEDS
FUEL PARTICLES
GAS COOLED REACTORS
GRAIN SIZE
HEAT TREATMENTS
HFIR REACTOR
HTGR TYPE REACTORS
IRRADIATION
IRRADIATION CAPSULES
JAEA
MICROSPHERES
NUCLEAR ENERGY
ORNL
POST-IRRADIATION EXAMINATION
SILICON CARBIDES
ZIRCONIUM CARBIDES