Loca analysis for assessment of stagnation feeder break size for taps#3&4

Abstract

During certain breaks that occur in the inlet piping connected to the fuel channel (single channel event) of horizontal core of Pressurized Heavy Water Reactor (PHWR), fuel damage in affected channel can not be prevented. The affected channel experiences sustained very low flow condition during a narrow range of breaks in inlet piping/feeder known as stagnation feeder break. In this type of accident, flow through the affected channel gets stopped or reduces to very low value when reactor is operating at full power causing a gross mismatch of the heat generation and heat removal in the channel, thus leading to rapid temperature rise in the affected channel. A study has been carried out to examine this aspect for the range of break size at different locations for Tarapur Atomic Power Station-3&4 (540 MWe). This paper brings out the assessment carried out for this type of accident.