Coupled calculations for the fast reactors safety justification with the EUCLID/V1 integrated computer code

The EUCLID/V1 integrated computer code is designed for the safety analysis and justification of the NPPs with liquid metal cooled fast reactors under normal operating conditions, normal operation failure, design basis accidents and beyond design basis accidents. The EUCLID/V1 code includes the system thermohydraulic module, spatial time-dependent neutronic modules, quasi two-dimensional fuel rod module and the modules of burnup and decay heat calculations. In the neutronic modules the improved quasistatic method is employed to solve the transport equation in the multigroup diffusion or discrete ordinates (Sn) forms. This work is devoted to the simulation results of some of the BREST-OD-300 and BN-1200 normal operation failure and accidents by means of the EUCLID/V1 code. Particularly, the loss of offsite power regime in the BN-1200 reactor has been modeled. In this case the reactor shutdown cooling with the emergency heat removal system is considered. It has been shown that after the control rods drop and reactor pumps shutdown the efficiency of two of four channels of the emergency heat removal system is sufficient to prevent the maximum cladding temperature from exceeding the safety limits. In frame of the test calculations the accident caused by the insertion of the total positive reactivity margin via withdrawal of all control rods from the reactor core at the maximum design speed during full power operation without scram operation (UTOP+ULOF) has been simulated for the BREST-OD-300 reactor facility. The obtained results indicate that after the reactor coolant pumps shutdown the total power decreases due to the thermohydraulic reactivity feedbacks and operation of the passive feedback system. By 100 s of the scenario natural circulation of lead progresses in the primary side, it is equal to 9.6% of the nominal total flow. The core disruption does not happen. (author)