Thermal-hydraulic analysis of main steam line breaks with continuous feedwater addition

Results are presented from a matrix of thermal-hydraulic analyses of postulated main steam line breaks for the Indian Point 3 Nuclear Power Plant. The matrix of cases includes various initial power levels, break sizes and types with the assumption of a feedwater closure valve failure. The feedwater flow used models of the pump characteristics, runout conditions and generator pressure dependence. The analyses performed with the RETRAN02 code provide primary coolant temperature and pressure, core reactivity, and break blowdown flowrate time histories for containment pressure response calculations