Evaluation of Sokrat Code Possibility to Model Uranium-Dioxide Fuel Dissolution by Molten Zirconium

K. S. Dolganov; A. E. Kiselev; N. I. Ryzhov; D. Yu. Tomashchik; F. Filippov; R. V. Chalyi; T. A. Yudina; S. A. Shevchenko; D. A. Yashnikov; N. A. Kozlova

doi:10.1007/s10512-018-0446-x

Evaluation of Sokrat Code Possibility to Model Uranium-Dioxide Fuel Dissolution by Molten Zirconium

Autores: Dolganov K.S.¹, Kiselev A.E.¹, Ryzhov N.I.¹, Tomashchik D.Y.¹, Filippov F.¹, Chalyi R.V.¹, Yudina T.A.¹, Shevchenko S.A.², Yashnikov D.A.², Kozlova N.A.²
Afiliações:
1. Nuclear Safety Institute, Russian Academy of Sciences (IBRAE RAN)
2. Scientific and Engineering Center for Nuclear and Radiation Safety (NTTs YaRB)
Edição: Volume 125, Nº 2 (2018)
Páginas: 82-90
Seção: Article
URL: https://journals.rcsi.science/1063-4258/article/view/249041
DOI: https://doi.org/10.1007/s10512-018-0446-x
ID: 249041

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A quantitative assessment is made of the possibilities of the SOKRAT code to model the dissolution of uranium dioxide fuel by zirconium cladding melt at the initial stage of a serious accident at NPP with VVER. The methodological approach for the assessment is based on the ASME V&V 20 standard and includes an uncertainty analysis. The results of local high-temperature experiments studying the kinetics of the process are used as a technical base.