Thermal Analysis of Severe Channel Damage Caused by a Stagnation Channel Break in a PHWR

Abstract

The reactor channel of the horizontal core of pressurized heavy water reactors experiences very low sustained flow during loss of coolant accident (LOCA) at the reactor inlet feeders caused by certain breaks known as critical channel breaks. In this type of accident the reactor trip is delayed 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 identify the phenomena and the break size leading to such a situation. Severe fuel damage is predicted in the channel.

Simulation of pressure tube deformation during high temperature transients

“…A typical break in the reactor inlet feeder produces flow stagnation in the reactor channel as reported by Mukhopadhyay et al [10]. During such a condition, it is found that the PT experiences a high temperature ramp of Fig.…”

“…Kundurpi and Luxat [6] and Oh et al [8] reported an early ballooning for a postulated case of 35% pump discharge line break and 30% reactor inlet header break for CANDU type reactors. A similar postulated case was analyzed by Mukhopadhyay et al [10] for a stagnation channel break caused by a single feeder pipe failure in Indian PHWRs. These cases are categorized as single failure events.…”