期刊
JOURNAL OF NUCLEAR MATERIALS
卷 491, 期 -, 页码 126-137出版社
ELSEVIER SCIENCE BV
DOI: 10.1016/j.jnucmat.2017.04.030
关键词
Molten (liquid) corium; Molten (liquid) UO2; Molten (liquid) ZrO2; Classical molecular dynamics; Density; Specific heat; Heat of fusion; Compressibility; Viscosity; Surface tension; Thermal conductivity
资金
- NSF program on Thermal Transport Processes [CBET1332807]
- National Nuclear R&D Program through the National Research Foundation of Korea (NRF) - Ministry of Science, ICT and Future Planning [2013M2B2A9A03051257]
- Basic Science Research Program through NRF - Ministry of Education [2015R1D1A1A01059621]
- National Research Foundation of Korea [2015R1D1A1A01059621, 2013M2B2A9A03051257] Funding Source: Korea Institute of Science & Technology Information (KISTI), National Science & Technology Information Service (NTIS)
Predicting the fate of accident-melted nuclear fuel-cladding requires the understanding of the thermophysical properties which are lacking or have large scatter due to high-temperature experimental challenges. Using equilibrium classical molecular dynamics (MD), we predict the properties of melted UO2 and ZrO2 and compare them with the available experimental data and the predictive models. The existing interatomic potential models have been developed mainly for the polymorphic solid phases of these oxides, so they cannot be used to predict all the properties accurately. We compare and decipher the distinctions of those MD predictions using the specific property-related autocorrelation decays. The predicted properties are density, specific heat, heat of fusion, compressibility, viscosity, surface tension, and the molecular and electronic thermal conductivities. After the comparisons, we provide readily usable temperature-dependent correlations (including UO2-ZrO2 compounds, i.e. corium melt). (C) 2017 Elsevier B.V. All rights reserved.
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