| Article ID | Journal | Published Year | Pages | File Type |
|---|---|---|---|---|
| 297244 | Nuclear Engineering and Design | 2012 | 14 Pages |
Lattice codes are able to produce burnup dependent fuel assembly homogenized neutron cross sections and pin power form factors. These data are used by diffusion nodal codes to produce fuel assembly powers and intra assembly power distributions. 2D transport codes are usually used for this type of the calculation. Four lattice codes (DRAGON 4, FA2D, NEWT and PSG2/Serpent) were used to calculate pin power distribution for reflected single fuel assemblies during depletion. The NPP Krško 16 × 16 fuel assemblies without and with IFBAs (Integral Fuel Burnable Absorber), for two U-235 enrichments 4.40% and 4.80%, as found in cycle 24 of the plant, were modeled. The heterogeneous depletion calculation was performed up to burnup of 60 GWd/tU. The prediction capabilities of the codes are compared for infinite multiplication factor, some group cross sections, and for 2D pin power form factors, in order to assess their applicability for preparing cross section data for diffusion nodal codes.
► We model fuel assembly depletion by DRAGON, FA2D, NEWT, and Serpent lattice codes. ► Satisfactory multiplication factor prediction with moderate influence of burnup. ► Prediction of maximum pin power within 2% with only a small burnup dependence. ► Maximum relative differences in 2D pin powers within 5%, no significant burnup influence. ► Up to 5% lower maximum pin powers obtained for pin-by-pin compared to uniform composition depletion.
