The AP600 reactor core approaches buoyancy-dominated flow at the departure from nucleate boiling (DNB)-limiting period of a postulated steam-line-break accident. The reactor core has a highly skewed power distribution at this time due to the conservative assumption of a withdrawn rod cluster control assembly (stuck rod). Under such conditions, strong buoyancy-induced core cross flow occurs, and coupled nuclear and thermal-hydraulic interactions become important. To analyze the transient, Westinghouse Electric Corporation has coupled THINC-IV with a neutronic code (ANC). Applicability of the THINC-IV subchannel code to the low-flow conditions with a steep radial power gradient is verified with existing rod bundle test results. The code predictions are in excellent agreement with the test data. The coupled codes provide a realistic three-dimensional simulation of core power by considering core flow distributions and the resultant enthalpy distributions in neutronic feedback. The safety analysis using the coupled code demonstrates that the DNB design basis is met during the postulated steam-line-break accident.