The steam line break (SLB) accident in pressurized water reactors is characterized by a large asymmetric cooling of the core, asymmetric stuck control rods, and large primary coolant flow variations. Because of these space- and time-dependent neutronic and thermal-hydraulic conditions in the core, former SLB analyses that used simplified core models were usually performed with many conservative assumptions. To clarify the complicated behavior of the core, the three-dimensional neutronic code CRONOS-2, the three-dimensional core thermal-hydraulic code FLICA-4, and the system code FLICA-S are completely coupled. The results obtained from the coupled codes indicate that the local thermal-hydraulic feedback effects are important in mitigating neutronic power excursions during SLBs.