The Shutdown Heat Removal Tests (SHRT), including the protected loss-of-flow test SHRT-17 and the unprotected loss-of-flow test SHRT-45R, have been performed at the EBR-II reactor, in order to confirm the inherent safety character and robustness of this facility. Since primary coolant temperatures, flowrates and static pressures were measured at various locations during these two tests, computational tools, being able to accomplish static and transient analysis of Sodium-cooled Fast Reactor (SFR), can be validated by benchmarking with these measured data. In the current study, TRACE code was used for transient simulations based on core performance parameters and safety related coefficients calculated by the Serpent Monte Carlo code. The TRACE code we used for this study has been extended by the FAST group in Paul Scherrer Institut (PSI) to achieve proper thermal-hydraulic assessments of Generation IV fast reactor systems cooled by liquid sodium. Core performance parameters and safety coefficients calculated with our Serpent model were proved to be accordant with the ones provided by ANL by using the DOT-III deterministic code, which confirms the simplifications we made during the Serpent model creation and also the reliability of the Serpent Monte Carlo code for the metallic fuel loaded sodium cooled fast reactor. Moreover, based on comparisons of core inlet/outlet coolant temperatures, IHX primary inlet temperatures, IHX secondary outlet temperatures and primary coolant flowrates with the measured data provided by ANL, our TRACE model was also proved to be able to provide satisfying results. These transient simulation results also confirm the eligibility of the EBR-II core to survive both the SHRT-17 and SHRT-45R type transients.