Abstract: Experimental Breeder Reactor Ⅱ (EBR-Ⅱ) of Argonne National Laboratory (ANL) is a sodium cooled fast reactor. The SHRT45R unprotected loss of flow test which demonstrated the effectiveness of passive feedback in the EBRⅡ reactor was carried out on April 3, 1986. The SHRT45R test was analyzed using the FASYS code which was developed at China Institute of Atomic Energy for thermal, hydraulic, and neutronic analysis of system transients in sodiumcooled fast reactor. The FASYS code contains point reactor model, decay heat model, reactivity feedback model, singlepin model for fuelpin temperature calculation, core and circuit hydraulic model, intermediate heat exchanger model, pump model, decay heat exchanger model and sodiumtoair heat exchanger model, etc. The core and primary circuit of the SHRT45R test were modeled by the FASYS code. Besides, the primary pump speed, the intermediate heat exchanger inlet flow and inlet temperature were taken as the calculation boundary conditions. The calculation was divided into two stages. In the initial simulation stage, the reactivity feedback was not calculated, the core power was taken as the input data, and only the thermal hydraulic behaviors of the core and the primary circuit of SHRT45R were calculated. In the final simulation stage, the reactivity feedback, the core power, and the thermal and hydraulic behavior of the core and the primary circuit were calculated simultaneously. The measured data were compared with calculated results of the FASYS code. It shows that the pump flowrate and XX09 instrumented subassembly coolant temperature predicted by the FASYS code are in good agreement with the measured data when the core power is taken as the input data. The Zpipe inlet temperature is predicted faster due to modeling the upper plenum as a 0D volume. Since the influence of adjacent hightemperature subassemblies around the lowtemperature XX10 instrumented subassembly is not considered, the coolant temperature of XX10 instrumented subassembly predicted is significantly lower than the measured data. In the final simulation of core power and temperature coupled calculation, the calculated value of core power is in good agreement with measured power overall. Specifically, the total core power by the FASYS code is slightly underpredicted between 1 and 4 minutes, and slightly overpredicted after 6 minutes. The increase in total core power after 10 minutes is due to the auxiliary EM pump head increasing, which leads to lower core temperatures and a small reactivity increase. The core thermal hydraulic model, primary circuit thermal hydraulic model, point reactor model, especially the reactivity feedback model of the FASYS code were verified by analysis of SHRT45R test. When the core outlet temperature changes rapidly, the temperature change trend of core outlet plenum simulated by the 0D volume model will be faster than the actual situation. In the future, the thermal stratification model should be considered for core outlet plenum temperature simulation and the subchannel code should be used for the center lowpower subassembly and the surrounding highpower subassemblies transient thermal simulation.