Basic verification of THACS for sodium-cooled fast reactor system analysis

For the transient thermal-hydraulic analysis of sodium cooled fast reactor (SFR), a universal code THACS is being developed in China. The one-dimensional modules for the main components of the primary loop, intermediate loop and heat removal loop have been compiled and the models are introduced briefly in this paper. For two-phase calculation, the multi-bubble model is used in the core module for sodium, while a simple incompressible water model is applied in the water side of steam generator. The point kinetic equations are employed to calculate the fission power of the multi-channel core, and various kinds of feedbacks like Doppler Effect, radial and axial expansions, density effect are taken into account. Preliminary verifications of the code are done based on the benchmark analyses of a BN-800 type SFR under loss of flow (LOF) accident, the EBR-II shutdown heat removal tests, and the 1-D simulation of the out-of-pile LOF experiment in KNS. Code-to-code and code-to-experiment comparisons show that THACS is proper for the transient analysis of SFR.