TRACE Analysis of 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. In the present study, based on FIX- II LOCA tests, simulation models for the tests of No. 3025, No. 3061 and No. 5052 were developed to validate the TRACE code (version 5.0 patch 2). The simulated transient thermal-hydraulic behaviors during the LOCA tests including the pressure in the primary system, the mass flow rate in certain key parts, and the temperature in the core are compared with experimental data. The simulation 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, choked flow model , insulator in the steam dome, K-factor in the test section, and pump trip, on the results. The sensitivity analyses show that both the models and parameters have significant influence on the outcome of the model.
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