| Article ID | Journal | Published Year | Pages | File Type |
|---|---|---|---|---|
| 298184 | Nuclear Engineering and Design | 2010 | 10 Pages |
Inspections of existing nuclear power plants have pointed out the possibility that small break loss-of-coolant accidents (SBLOCAs) could be initiated by a small break located in the upper head (UH) of the reactor pressure vessel (RPV). Such type of breaks has been the subject of investigation in some of the tests carried out in the framework of the OECD/NEA ROSA test program for safety research and safety assessment of light water reactors. The ROSA/LSTF test facility simulates a Westinghouse design PWR with a four-loop configuration and 3423 MWth. Areas, volumes and power are scaled down by a factor of 1:48 while the elevations are kept at full height. Only two loops, sized to conserve the volume scaling (2:48), are simulated. The present paper is focused on test 6-1 that simulated a RPV upper head SBLOCA with a break size equivalent to 1.9% cold leg break. The experiment assumes a total failure of the high pressure injection system (HPIS) and a loss of off-site power concurrent with the scram. The main purpose of the present study is the assessment of the capabilities of the best estimate system code, TRACE, to reproduce and understand the physical phenomena involved in this type of SBLOCA scenarios. Special attention was dedicated to the modelling of the leakage flows, necessary to correctly simulate the distribution of the water inventory in the primary side. In addition, the particular location of the break in test 6-1 allows the verification of the chocked flow model in the same way as for a separate-effect test.
