# The MCU-REA Code

Abagjan L.P., Alekseev N.I., Bryzgalov V.I., Glushkov A.E., Gomin E.A., Gurevich M.I., Kalugin M.A., Majorov L.V., Marin S.V., Yudkevich M.S.

The MCU-REA code allows solving the neutron transport equation using the Monte-Carlo method and evaluated nuclear data. In particular, it calculates the criticality parameters and solves burnup problems for VVER type systems.

The possibility to calculate neutron characteristics of the core taking into account changes in the isotopic composition of the fuel is the main peculiarity of the MCU-REA code.

Main application fields:

- criticality safety assessment,
- calculation of neutron physical characteristics of VVER type reactors,
- verification and validation of nuclear data libraries,
- verification of design codes intended for calculation of neutron physical characteristics of VVERs.

The code allows one to take into account the effects of continuous change of the energy at collisions, and both pointwise and step representation of cross-sections. For the unresolved resonance region the cross-sections are calculated using the temperature dependent subgroup parameters or the Bondarenko’s self-shielding f-factors. For the resolved resonance region both subgroup and pointwise descriptions of cross-sections are possible. In latter case, “infinite” number of points describes the cross-sections of the most important nuclides. They are calculated in each energy point using the resonance parameters at neutron history modeling. This scheme allows one to estimate temperature effects using the analytical dependence of cross-sections on temperature. Collision modeling in the thermalization region is performed according to the user’s choice: in multi-group approximation, or using the model of continuous energy change taking into account correlation between energy and angle change at scattering. The both cases use S(a ,b ) formalism and take into account chemical binding, crystal structure, and nucleus thermal motion.

The MCU-REA code allows the user to model practically any three-dimensional systems described by means of combinatorial geometry method. The user may describe lattices with repeated elements defining the repeated elements and translation vectors. The lattices may include heterogeneity described as applications. The combinatorial approach is strengthened by the use of the Woodcock method. Double heterogeneous systems with fuel elements containing many thousands of sphere microcells may be modeled using a special algorithm.

Different flux functionals are calculated: neutron multiplication factor, effective fraction of delayed neutrons, nuclear reaction rates for separate nuclides and their mixture in the given space-energy intervals, few-group constant set for regions, cells and assemblies (including diffusion coefficients).