TY - JOUR
T1 - In situ ion irradiation of zirconium carbide
AU - Ulmer, Christopher J.
AU - Motta, Arthur T.
AU - Kirk, Mark A.
N1 - Funding Information:
This work was funded by the U.S. Department of Energy's Nuclear Engineering University Program project number 10-679 . The electron microscopy with in situ ion irradiation was accomplished at Argonne National Laboratory at the IVEM-Tandem Facility, a U.S. Department of Energy Facility funded by the DOE Office of Nuclear Energy , operated under Contract No. DE-AC02-06CH11357 by UChicago Argonne, LLC. We thank Pete Baldo and Ed Ryan of Argonne National Laboratory for their invaluable assistance in carrying out the irradiations. We also thank Ming-Jie Zheng, Izabela Szlufarska, Dane Morgan and Yina Huang for their insights and discussions.
Publisher Copyright:
© 2015 Elsevier B.V.
PY - 2015/11/1
Y1 - 2015/11/1
N2 - Zirconium carbide (ZrC) is a candidate material for use in one of the layers of TRISO coated fuel particles to be used in the Generation IV high-temperature, gas-cooled reactor, and thus it is necessary to study the effects of radiation damage on its structure. The microstructural evolution of ZrCx under irradiation was studied in situ using the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory. Samples of nominal stoichiometries ZrC0.8 and ZrC0.9 were irradiated in situ using 1 MeV Kr2+ ions at various irradiation temperatures (T = 20 K-1073 K). In situ experiments made it possible to continuously follow the evolution of the microstructure during irradiation using diffraction contrast imaging. Images and diffraction patterns were systematically recorded at selected dose points. After a threshold dose during irradiations conducted at room temperature and below, black-dot defects were observed which accumulated until saturation. Once created, the defect clusters did not move or get destroyed during irradiation so that at the final dose the low temperature microstructure consisted only of a saturation density of small defect clusters. No long-range migration of the visible defects or dynamic defect creation and elimination were observed during irradiation, but some coarsening of the microstructure with the formation of dislocation loops was observed at higher temperatures. The irradiated microstructure was found to be only weakly dependent on the stoichiometry.
AB - Zirconium carbide (ZrC) is a candidate material for use in one of the layers of TRISO coated fuel particles to be used in the Generation IV high-temperature, gas-cooled reactor, and thus it is necessary to study the effects of radiation damage on its structure. The microstructural evolution of ZrCx under irradiation was studied in situ using the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory. Samples of nominal stoichiometries ZrC0.8 and ZrC0.9 were irradiated in situ using 1 MeV Kr2+ ions at various irradiation temperatures (T = 20 K-1073 K). In situ experiments made it possible to continuously follow the evolution of the microstructure during irradiation using diffraction contrast imaging. Images and diffraction patterns were systematically recorded at selected dose points. After a threshold dose during irradiations conducted at room temperature and below, black-dot defects were observed which accumulated until saturation. Once created, the defect clusters did not move or get destroyed during irradiation so that at the final dose the low temperature microstructure consisted only of a saturation density of small defect clusters. No long-range migration of the visible defects or dynamic defect creation and elimination were observed during irradiation, but some coarsening of the microstructure with the formation of dislocation loops was observed at higher temperatures. The irradiated microstructure was found to be only weakly dependent on the stoichiometry.
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U2 - 10.1016/j.jnucmat.2015.08.009
DO - 10.1016/j.jnucmat.2015.08.009
M3 - Article
AN - SCOPUS:84941796790
VL - 466
SP - 606
EP - 614
JO - Journal of Nuclear Materials
JF - Journal of Nuclear Materials
SN - 0022-3115
M1 - 49263
ER -