Approximation of WWER 440/213 neutron flux density as a function of the linear power and fuel burn-up

Fast neutron flux density, thermal neutron flux density or their ratio are usually needed for input into the fuel thermo-mechanical integral codes. The fast neutron flux has significant impact on cladding creep, irradiation growth and basic cladding mechanical properties. These properties are of special importance at high burn-up. The aim of the presented work was the approximation of the neutron flux density as a function of local power and burn-up for various neutron energy intervals. The input data were presented by the computer code MOBY-DICK during Dukovany NPP 4-year fuel cycle simulation. It was the time-space distribution of neutron flux densities in fuel element with energies E < 0,625 eV (thermal neutrons) and E > 1 MeV (fast neutrons), resp. The presented fits should be taken as a first rough attempt to open the unresolved problem. More detailed analyses from the neutron-physics side using specialized tools are needed.(Authors)