Abstract
The ROSA (Rig of Safety Assessment)-III facility is a volumetrically scaled (1/424) simulated boiling water nuclear reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. A recirculation pump suction line break test with a five percent break area was conducted with the assumption of high pressure core spray system (HPCS) failure. The simulated peripheral fuel rods facing the channel box wall had a tendency to be rewetted temporarily at the upper part of the core by falling water from the upper plenum before low pressure core spray system (LPCS) actuation, while the rods in the central region were not rewetted but quenched mainly from the bottom of the core after low pressure coolant injection system (LPCI) actuation. Therefore, the peak cladding temperatures of the simulated high power fuel rods were limited to lower values since they were located in the peripheral region and the temporary rewetting before LPCS actuation occurred mainly in the peripheral region. The ROSA-III five percent break test and a BWR counterpart were analyzed with the RELAP5/MOD1 (cycle 018) code. Similarity between the ROSA-III small break test and a BWR small break LOCA has been confirmed through comparison of the calculated results.
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