CONCEPTUAL DESIGN STATION BLACKOUT AND LOSS-OF-FLOW ACCIDENT ANALYSISFOR THE ADVANCED NEUTRON SOURCE REACTOR

A system model of the Advanced Neutron Source Reactor (ANSR) has been developed and used to perform conceptual safety analyses. To better re present thermal-hydraulic behavior in the unique geometry and conditio ns of the ANSR core, three specific changes in the RELAP5/MOD3 compute r code were implemented: a turbulent forced-convection heat transfer c orrelation, a critical heat flux correlation, and an interfacial drag correlation. The system model includes representations of the ANSR cor e, heat exchanger coolant loops, and the pressurizing and letdown syst ems. Analyses of ANSR station blackout and loss-of-flow accident scena rios are described. The results show that the core can survive without exceeding the flow excursion or critical heat flux thermal limits def ined for the conceptual safety analysis, if the proper mitigation opti ons are provided.