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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… Click to show full abstract
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.
Keywords:
dissolution;
uranium dioxide;
dioxide fuel;
model;
sokrat code
Journal Title: Atomic Energy
Year Published: 2018
Link to full text (if available)
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Journal Title: Atomic Energy
Year Published: 2018
Link to full text (if available)
Share on Social Media: Sign Up to like & get
recommendations!