Abstract
An advanced version of the BERKUT-U/V2.1 code intended for modeling the behavior of fuel rods with oxide or nitride fuel in standard and emergency operating regimes of fast reactors with liquid-metal coolant is being developed at the Nuclear Safety Institute, Russian Academy of Sciences (IBRAE RAS), as part of the project Next-Generation Codes of Project Proryv (Breakthrough). The present work reports the results of the validation of an advanced version of the code based on post-reactor studies of BN-600 fuel rods (KETVS-1, -6) and BREST-OD-300 fuel rods (ETVS-5) with mixed uranium-plutonium nitride fuel, irradiated in the BN-600 reactor.
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