A system model of the Advanced Neutron Source Reactor (ANSR) has been developed and used to perform conceptual safety analyses. To better represent thermal-hydraulic behavior in the unique geometry and conditions of the ANSR core, three specific changes in the RELAP5/MOD3 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux correlation, and an interfacial drag correlation. The system model includes representations of the ANSR core, heat exchanger coolant loops, and the pressurizing and letdown systems. Analyses of ANSR station blackout and loss-of-flow accident scenarios are described. The results show that the core can survive without exceeding the flow excursion or critical heat flux thermal limits defined for the conceptual safety analysis, if the proper mitigation options are provided.