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In this work was carried out the simulation in the SCALE6 code of an experiment on the BN-600 reactor on irradiating of fuel assemblies, containing samples of mixed nitride uranium-plutonium fuel. A comparison of the results for SCALE6 on the results of other codes is presented. The results of an estimation of uncertainties in the calculated data connected with uncertainties of an irradiation of an experimental sample and used neutron cross-sections library. Discussion of possible differences between analytical and experimental results is given.

Modeling of Analytical-Experimental Benchmark on Irradiation of Nitride Fuel in BN-600 Reactor in the SCALE6 code