Abstract: The shutdown heat removal test SHRT-17 was carried out in the Experimental Breeder Reactor Ⅱ (EBR-Ⅱ) in 1986. EBR-Ⅱ SHRT-17 test is protected loss of flow test. The purpose of SHRT-17 test is to prove that the natural circulation flow of EBR-Ⅱ can ensure the safety of the core when the primary circuit loses all forced circulation flow. The thermal-hydraulic analysis of EBR-Ⅱ SHRT-17 test was calculated by the fast reactor system analysis code FASYS which was developed by China Institute of Atomic Energy. The FASYS code has the ability to analyze transient responses of core, primary coolant system, secondary coolant system, and decay heat removal system in sodium-cooled fast reactor. The FASYS code has been verified by analytical solution, code-to-code validation, China Experimental Fast Reactor test data, and EBR-Ⅱ SHRT-45R test data. The EBR-Ⅱ primary sodium system was represented by 36 liquid segments in the FASYS model. And the 24-channel core model was used for the EBR-Ⅱ SHRT-17 test analysis. Besides, the core power, the primary pump speed, the intermediate heat exchanger intermediate inlet sodium mass flow rate and inlet temperature were taken as the calculation boundary conditions. The measured data of primary circuit flow rate and temperature were compared with calculated results of the FASYS code. It shows that No.2 primary pump flowrate, XX09 instrumented subassembly flowrate and coolant temperature predicted by the FASYS code are in good agreement with the measured data at the pump coasting stage and natural circulation stable stage. When the primary pump just stops coasting, the natural circulation flow in primary circuit is not stable, and the calculated value of the natural circulation flow by the FASYS code is higher than the measured data. In the natural circulation unstable stage, the core flow rate is very small, and the temperature of sodium fluid in core outlet plenum has a great influence on the core outlet temperature measurements. Since the natural circulation flow is overpredicted and the effect of sodium fluid in core outlet plenum is not considered, the core outlet temperature predicted drops faster than the measured data at natural circulation unstable stage. The thermal hydraulic behavior of the low-temperature XX10 instrumented subassembly was analyzed by the single-channel model, without considering the influence of adjacent high-temperature subassemblies. Therefore, the transient coolant temperature of the XX10 instrumented subassembly is underpredicted, and the predicted natural circulation driving force and flow is also lower than the measured data. The primary circuit thermal hydraulic model especially the core thermal hydraulic model of the FASYS code have been verified by analysis of SHRT-17 test.