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Atomic Energy

, Volume 125, Issue 2, pp 82–90 | Cite as

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
Article
  • 1 Downloads

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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Copyright information

© Springer Science+Business Media, LLC, part of Springer Nature 2018

Authors and Affiliations

  • K. S. Dolganov
    • 1
  • A. E. Kiselev
    • 1
  • N. I. Ryzhov
    • 1
  • D. Yu. Tomashchik
    • 1
  • F. Filippov
    • 1
  • R. V. Chalyi
    • 1
  • T. A. Yudina
    • 1
  • S. A. Shevchenko
    • 2
  • D. A. Yashnikov
    • 2
  • N. A. Kozlova
    • 2
  1. 1.Nuclear Safety InstituteRussian Academy of Sciences (IBRAE RAN)MoscowRussia
  2. 2.Scientific and Engineering Center for Nuclear and Radiation Safety (NTTs YaRB)MoscowRussia

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