Thermophysical properties of U, Zr-oxides as prototypic corium materials

JOURNAL OF NUCLEAR MATERIALS

卷:520

Article

Seibert, Alice¹  Staicu, Dragos¹  Bottomley, David¹  Cologna, Marco¹  Boshoven, Jacobus¹  Hein, Herwin¹  Kassim, Emthetal¹  Nourry, Sarah¹  Ernstberger, Markus¹  Robba, Davide¹  Konings, Rudy¹ 

[1] European Commiss, Joint Res Ctr, POB 2340, D-76125 Karlsruhe, Germany

关键词: Corium;    Thermal diffusivity;    Thermal conductivity;    ZrO2;    UO2;    Mixed oxides;    SPS;   

DOI  :  10.1016/j.jnucmat.2019.04.019

来源: Elsevier

【 摘 要 】

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-Zroxide 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. (C) 2019 The Authors. Published by Elsevier B.V.

【 授权许可】

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