Thermal-hydraulic transient analysis of the FFTF LOFWOS Test #13

Abstract

In the framework of a Coordinate Research Project (CRP) endorsed by the International Atomic Energy Agency (IAEA), the Department of Astronautical, Electrical and Energy Engineering (DIAEE) of Sapienza University of Rome has developed a thermal–hydraulic model of the Fast Flux Test Facility (FFTF). FFTF was a Sodium-cooled Fast Reactor (SFR) developed by the Westinghouse Electric Corporation for the U.S. Department of Energy, which operated from 1980 to 1992. The main mission of the reactor was to prove feasibility and operation of SFR nuclear power plants. For this purpose, FFTF was designed as a flexible system and extensively instrumented, making it attractive for code validation purposes. The proposed benchmark exercise involves a Loss Of Flow Without Scram (LOFWOS) test, performed in 1986. DIAEE has developed and is validating a multiphysics modelling based on a Neutron-Kinetic/Thermal-Hydraulic (NK/TH) coupling approach with RELAP5-3D© and PHISICS codes. The present paper deals with the assessment of the standalone TH modelling, establishing the basis for the next NK/TH phase. Steady-state results have been compared with benchmark specifications, proving the adequacy of the model calibration. After that, the comparison between transient simulation and experimental data has highlighted an overall good agreement, demonstrating the capability of the developed TH nodalization to reproduce the FFTF LOFWOS Test #13. Outside of the parameters selected for the comparison in the benchmark exercise, outlet plenum thermal stratification and sodium free surface motion within Gas Expansion Modules (GEM) have been analysed, providing useful outcomes for the understanding of the transient event and for future perspectives.