SPECTRA Code Validation through PWR Cold Leg Small Break LOCA Tests in OECD/NEA ROSA Project

摘要

OECD/NEA ROSA Test Series-5 experiments were conducted using the Large Scale Test Facility (LSTF) at the Japan Atomic Energy Agency. They were focused on the observation of two-phase natural circulation and non-uniform flow among the U-tubes. Both phenomena may appear during the primary cooling through the Steam Generator secondary-side depressurization as an accident management action with (test 5-2) and without (test 5-1) non-condensable gas inflow from the accumulator tanks. Both tests simulated a PWR 0.5% cold leg SBLOCA under the assumption of a total failure of high and low pressure injection systems. The objectives of the experiments were to obtain detailed thermal-hydraulic data suitable for the validation of computer codes such as SPECTRA. SPECTRA (Sophisticated Plant Evaluation Code for Thermal-hydraulic Response Assessment) is NRG in-house developed code, designed for thermal-hydraulic analyses of nuclear or conventional power plants. A model of the LSTF facility was developed in SPECTRA. The model focused on detailed simulation of the break flow and liquid level in different sets of Steam Generator U-tubes. The SPECTRA model predicted reasonably well the overall thermal-hydraulic phenomena observed in both experiments, although improvements could be deployed for a better agreement with experimental results of liquid level in the Steam Generator U-tubes in the case that non-condensable gas is present in the primary system.