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.
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Translated from Atomnaya Énergiya, Vol. 125, No. 2, pp. 79–86, August, 2018.
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Dolganov, K.S., Kiselev, A.E., Ryzhov, N.I. et al. Evaluation of Sokrat Code Possibility to Model Uranium-Dioxide Fuel Dissolution by Molten Zirconium. At Energy 125, 82–90 (2018). https://doi.org/10.1007/s10512-018-0446-x
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DOI: https://doi.org/10.1007/s10512-018-0446-x