Development of a subchannel analysis code for SFR wire-wrapped fuel assemblies

Abstract

Sodium cooled fast reactor (SFR) exhibit some unique thermo-hydraulic characteristics in comparison to thermal reactors, considering that SFR have a higher core power density, special configuration of assemblies, wire-wrapped spacers, special thermo-physical properties of sodium and so on. Some design limits related to fuel, cladding and coolant temperature must be met in both the normal and transient condition. Thus, it is of great importance to obtain accurate thermal-hydraulic performance for the design and safety assessment of the SFR reactor. Therefore, a subchannel analysis code SUBAC specially for SFR wire-wrapped assemblies is developed in the present study. Sensitivity analysis of different model combinations is carried out by using ORNL 19-pin benchmark test. The calculated results show that the normalized temperature at the exit of the rod bundle calculated by the Cheng-Todreas correlation combined with Cheng-Tak turbulent mixing model is consistent with the experimental data. Subsequently, the 19-pin and 37-pin bundle benchmark problems are calculated to validate the accuracy and applicability of the code. All the calculated results agree well with the experimental data, indicating that the code is competent for steady-state calculations. Finally, SUBAC code is used for performing transient analyses of EBR-II SHRT-17 XX09 subassembly. A good agreement between calculated and experimental values of the core top sodium temperature is achieved, but slightly underestimated. In conclusion, the SUBAS program is capable of being used for subchannel analysis of SFR and some suggestions for future research are given.