Preliminary Thermohydraulics Analysis of the AMPS-1000 Reactor Core Using COBRA-IV

Abstract

The critical power ratio (CPR) of the AMPS-1000 reactor was calculated using the subchannel code COBRA-IV. The code employs critical heat flux (CHF) correlations to predict the margin to critical heat flux. Three new CHF correlations have been implemented in the code, i.e, EPRI, look-up table, and pool boiling. The analysis was performed using the three correlations at different operating conditions. It was confirmed that the margin to CHF of the reactor is consistently high over the entire range of operating conditions.