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This paper describes the methods used in the Serpent 2 Monte Carlo code for producing homogenized group constants for nodal diffusion and other deterministic reactor simulator calculations. The methodology covers few-group reaction cross sections, scattering matrices, diffusion coefficients and poison cross sections homogenized in infinite and B1 leakage-corrected critical spectra, as well the calculation of assembly discontinuity factors, pin-power form factors, delayed neutron parameters and total and partial albedos. Also included is a description of an automated burnup sequence, which was recently implemented for the handling of restart calculations with branch variations. This capability enables covering the full range of local operating conditions required for the parameterization of group constants within a single run. The purpose of this paper is to bring the methodological description provided in earlier publications up to date, and provide insight into the developed methods and capabilities, including their limitations and known flaws.

This paper describes the methods used in the Serpent 2 Monte Carlo code for producing homogenized group constants for nodal diffusion and other deterministic reactor simulator calculations. The methodology covers few-group reaction cross sections, scattering matrices, diffusion coefficients and poison cross sections homogenized in infinite and B1 leakage-corrected critical spectra, as well the calculation of assembly discontinuity factors, pin-power form factors, delayed neutron parameters and total and partial albedos. Also included is a description of an automated burnup sequence, which was recently implemented for the handling of restart calculations with branch variations. This capability enables covering the full range of local operating conditions required for the parameterization of group constants within a single run. The purpose of this paper is to bring the methodological description provided in earlier publications up to date, and provide insight into the developed methods and capabilities, including their limitations and known flaws.

Subsequent studies have identified many scenarios which can lead to reactivity excursions due to boron dilution. The comparative study, presented in this paper, deals with the so-called "restart of the first reactor coolant pump" scenario and its reactor-dynamic consequences for the both VVER reactor types - VVER-440 and VVER-1000. The transient simulations were performed using the three-dimensional core dynamics code DYN3D. The DYN3D modeling features, including recent developments, as well as the cross-section generation methodology, involved in these calculations, are described. The analyzed accident scenario is outlined together with the assumptions made. The results of core response in this boron dilution accident for both VVER reactors have been compared within ranges, determined by the two reactivity values of interest: the criticaly limit and the reactivity initiated accident (RIA) limit.