SUBCHANNEL THERMAL-HYDRAULIC ANALYSIS AT AP600 LOW-FLOW STEAM-LINE-BREAK CONDITIONS

The AP600 reactor core approaches buoyancy-dominated flow at the depar ture from nucleate boiling (DNB)-limiting period of a postulated steam -line-break accident. The reactor core has a highly skewed power distr ibution at this time due to the conservative assumption of a withdrawn rod cluster control assembly (stuck rod). Under such conditions, stro ng buoyancy-induced core crossflow occurs, and coupled nuclear and the rmal-hydraulic interactions become important. To analyze the transient , Westinghouse Electric Corporation has coupled THING-IV with a neutro nic code (ANC). Applicability of the THING-IV subchannel code to the l ow-flow conditions with a steep radial power gradient is verified with existing rod bundle test results. The code predictions are in excelle nt agreement with the test data. The coupled codes provide a realistic three-dimensional simulation of core power by considering core flow d istributions and the resultant enthalpy distributions in neutronic fee dback. The safety analysis using the coupled code demonstrates that th e DNB design basis is met during the postulated steam-line-break accid ent.