TRACE Validation Against LOCA Transients Performed on FIX-II Facility

As a latest developed computational code, TRACE is expected to be useful and effective for analyzing the thermal-hydraulic behaviors in design, licensing and safety analysis of nuclear power plant. However, its validity and correctness have to be verified and qualified before its application into industry. Loss-of-coolant accident (LOCA) is a kind of transient thermal hydraulic event which has been emphasized a lot as a most important threat to the safety of the nuclear power plant. The FIX-II experiments were performed to produce experimental data for understanding the initial stage of LOCA and so as to verify the computational codes. In the present study, based on FIX-II LOCA tests, simulation models for the tests of No. 3025, No. 3061 and No. 5052 which correspond to different LOCA cases were developed to validate the TRACE code (version 5.0 patch 2). The predictions of the TRACE code including the pressure in the primary system, the mass flow rate in certain key parts, and the temperature in the core were compared with the experimental data. The results show that TRACE model can well reproduce the transient thermal-hydraulic behaviors under different LOCA situations. In addition, sensitivity analysis are also performed to investigate the influence of particular models and parameters, including counter current flow limitation (CCFL) model and choked flow model on the results, which show that both the models have significant influence on the outcome of the model.