Preliminary Thermalhydraulics Analysis of the AMPS-1000 Reactor Using COBRA-IV and RELAP5/MOD2

Abstract

The thermal hydraulic behaviour of the AMPS-1000 reactor is analyzed using the subchannel code COBRA-IV and the system code RELAP5/MOD2. COBRA-IV was used to calculate the Critical Power Ratio (CPR) and the Minimum Departure from Nucleate Boiling (MDNBR) of the core under normal and accident conditions, while RELAP was used to simul ate system thermal hydraulic response under different accident scenarios. Loss Of Flow, Loss Of Heat Sink, Loss Of Coolant and Loss Of Reactor Power Regulation accidents were simulated in this analysis. In order to evaluate the inherent safety features of the AMPS-1000 design, no shutdown system action was credited and the reactor core behaviour is governed by the reactivity feedbacks. The analysis demonstrates that the self-regulating features of the reactor are effective in controlling the power during these accident scenarios and the fuel remains well-cooled throughout the transient.