Kovalev N.V.1, Prokoshin A.M.1, Kudinov A.S.1, Nevinitsa V.A.2, Nikandrova M.V.1, Goletsky N.D.1
1 Joint-stock company “Khlopin Radium Institute”, St. Petersburg, Russia
2 National Research Centre “Kurchatov Institute”, Moscow, Russia
Cross-verification and validation of the world famous and widely used software implementing the Monte Carlo method in the field of determining the nuclide composition of spent nuclear fuel (SNF) have been carried out. SCALE 6.3, MCNP 6.1, Serpent 2.1.32 were used. Comparison of the calculation results was carried out with the experimentally studied nuclide composition of the SNF sample from the Balakovo-2 NPP, which burnup was 46.7±0.7 MW∙day/kgHA. The nuclide composition was calculated using the infinite fuel assembly model. In addition, the calculation was performed using the deterministic two-dimensional code SCALE (NEWT). The calculation results for all the developed models showed an error at the level of 10 % in comparison with the experiment. It should be noted by default the Serpent code does not take into account the probabilities of isomer yields, and therefore the concentration of the 236Pu isotope turns out to be several orders of magnitude lower than it should be. The problem is solved by connecting the library of isomer yield probabilities in the ENDF 6 format. Also, for some reason, in the used version of MCNP 6.1 Cloud, the probabilities of isomer yield in the 237Np (n, 2n) reaction are confused, as a result the 236Pu concentration is reduced by about five once. The 236Pu isotope is the parent of 232U with a half-life of 2.858 years. The decay chain of the 232U isotope has powerful gamma sources. As a result of incorrect calculation of the 236Pu isotope, the radiation characteristics of spent nuclear fuel will be underestimated. Taking into account the remarks, all considered software can be recommended for use in determining the compositions of the spent nuclear fuel of thermal reactors.
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