A postulated single control rod drop transient was calculated for a typical boiling water reactor plant taking into account effects of detailed void distributions in those bundles neighboring the withdrawn control blade. Time-dependent pin power distributions were reconstructed by the plant simulator TRAC/BF1-ENTRÉE and were exported to the subchannel code NASCA.
Macroscopic cross-section libraries based on flat and distorted void distributions were allocated in accordance with fuel location in a simplified two-way coupling method. Exposure trends of bundle neutronic properties were compared between two void distributions. Although the infinite multiplication factor was not influenced, the radial peaking factor increased significantly because of the void distortion caused by pin-by-pin exposure of fissile materials.
The result with the combined cross sections was compared with those with the flat void cross sections. Application of the combined cross sections lowered the initial local peaking because of larger neutron leakage around the withdrawn control blade. The transient linear power density at the critical fuel rod increased more rapidly. A change in the fuel heat flux was attenuated because of the heat conduction delay. As a consequence of these effects, the peak cladding temperature became slightly lower than that of the flat void model.
