Solving the criticality problem for selected variants of the fuel pin lattice by the MCNP code

The MCNP code, which is based on the Monte Carlo method, was applied to criticality calculations with respect to the neutron transport equation. Six variants of the lattice parameters were used. In parallel, the point cross section data using the ENDF/B-4, ENDF/B-5, and ENDF/B-6 libraries were calculated for each variant. The multiplication factors are tabulated. The table also includes data obtained by the MCU code. The multiplication factors obtained when using the ENDF/B-4 library are systematically lower than those obtained when using the ENDF/B-5 and ENDF/B-6 libraries, and it is concluded that the two latter libraries are superior to the former one. The applicability of the MCU code is discussed. (P.A.)