Abstract
Several experimental facilities, such as the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA), have been built to reproduce some accidental scenarios because full-scale testing is usually impossible to perform.
One of the objectives of these Integral Test Facilities (ITFs) is to obtain measured data to be compared to simulations in order to test the capability of the thermalhydraulic codes to reproduce experimental conditions.
The applicability of these experimental results to a full-size power plant system depends on the scaling criteria adopted.
The present paper is focused on the simulation and the scaling of the Test 1-2 in the frame of the OECD/NEA ROSA Project to a Nuclear Power Plant (NPP). This test simulates a hot leg 1% Small Break Loss-Of-Coolant Accident (SBLOCA) in a Pressurized Water Reactor (PWR) under the actuation of High Pressure Injection (HPI) system and Accumulator Injection System (AIS).
A scaled-up NPP TRACE5 input has been developed from a LSTF TRACE5 model validated by authors in previous works. The scaled-up model has been developed conserving the power-to-volume scaling ratios of LSTF components, initial and boundary conditions. Lengths and diameters of hot legs have been scaled from LSTF model trying to conserve Froude number.
A comparison between both TRACE5 models (LSTF and scaled-up NPP) is performed (system pressures, discharged inventory and collapsed liquid levels). Special TRACE5 models such as Choked flow model and OFFTAKE model have been tested. A 3D VESSEL component has been tested in comparison to 1D TEE component to simulate the hot leg where the SBLOCA is located and varying the break orientation (downwards and upwards). Finally, a sensitivity analysis has been made to determine the effect of the break size in the SBLOCA range.
