State-of-the-art computational method for safety evaluation of sodium-cooled fast reactor

Uchibori, Akihiro ; Aoyagi, Mitsuhiro ; Sonehara, Masateru ; Okano, Yasushi; Takata, Takashi*

The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. A new computational code, SPECTRA, for integrated analysis of in- and ex-vessel phenomena during severe accidents in sodium-cooled fast reactors is a key technology of ARKAIDA-Safety which aims for safety evaluation. The in-vessel thermal hydraulics module in this code includes coupled analytical models for multidimensional multifluid model considering compressibility and relocation of a molten core. A lumped mass model is employed for computing behavior of ex-vessel compressible multicomponent gas including aerosols. This model is coupled with the models for ex-vessel phenomena such as sodium fire. Loss of reactor level event starting from leakage of sodium coolant was computed. Basic capability to evaluate severe accident progress was demonstrated through this analysis.