In situ ion irradiation of zirconium carbide

JOURNAL OF NUCLEAR MATERIALS

卷:466

Article

Ulmer, Christopher J.¹  Motta, Arthur T.¹  Kirk, Mark A.² 

[1] Penn State Univ, Dept Mech & Nucl Engn, University Pk, PA 16802 USA

[2] Argonne Natl Lab, Nucl Engn Div, Argonne, IL 60439 USA

关键词: Nuclear materials;    Zirconium carbide;    Ion irradiation;    Defects;   

DOI  :  10.1016/j.jnucmat.2015.08.009

来源: Elsevier

【 摘 要 】

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. (C) 2015 Elsevier B.V. All rights reserved.

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