Title: Thermophysical properties of U, Zr-oxides as prototypic corium materials
Authors: SEIBERT ALICESTAICU DRAGOSBOTTOMLEY PAULCOLOGNA MARCOBOSHOVEN JACOBUSHEIN HERWINKASSIM EMTETHALSTOHR SARAHERNSTBERGER MARKUSROBBA DAVIDEKONINGS RUDY
Citation: JOURNAL OF NUCLEAR MATERIALS vol. 520 p. 165-177
Publisher: ELSEVIER SCIENCE BV
Publication Year: 2019
JRC N°: JRC105924
ISSN: 0022-3115 (online)
URI: http://publications.jrc.ec.europa.eu/repository/handle/JRC105924
DOI: 10.1016/j.jnucmat.2019.04.019
Type: Articles in periodicals and books
Abstract: Simulated corium samples were prepared using a sol-gel process to yield U-Zr-oxide materials representative of a molten core covering the whole range of compositions in the U-Zr series. Discs of U-Zr-oxide were compacted by Spark Plasma Sintering (SPS). The materials were characterised by XRD and optical/electron microscopy techniques as well as SEM-EDX. The thermal diffusivity of all samples has been measured between 500 and 1600 K by the laser-flash technique and thermal conductivity was calculated. For comparison, a sample extracted from the fully melted core of the Three Mile Island reactor Unit 2 (TMI-2) was also investigated. The results for the simulated and real corium were analysed and compared to literature data. A substantial decrease of the thermal diffusivity occurred as the fraction of ZrO2 increased up to 18 mol% in the simulated corium. In the range 18–74 mol% ZrO2 only a weak composition dependence was observed. In this range the thermal conductivity at 500 K is between 2.5 and 3 W m−1 K−1, in agreement with other experimental data.
JRC Directorate:Nuclear Safety and Security

Files in This Item:
There are no files associated with this item.


Items in repository are protected by copyright, with all rights reserved, unless otherwise indicated.