Creep-Fatigue Damage Assessment of Reactor Hot Pool Components During Crash Cooling

Abstract

In the prototype fast breeder reactor (PFBR), when there is a reactor SCRAM, the shutdown takes place in two phases. First, there is hot shutdown, i.e. the hot pool sodium temperature is rapidly brought down from 820 to 623 K in 25 min, and the II phase is to proceed to the cold shutdown (453 K isothermal) condition gradually by deploying operating grade decay heat removal system (OGDHR ). The controlled cooling involving high and varying operating mode of suction pressure of the pump makes the pump design complex. In view of this, the requirement of controlled cooling has been relooked in the commercial breeder reactor (CBR), and it is proposed to deploy the OGDHR only after reaching cold shutdown. Thermo-mechanical analysis of the reactor components, i.e. control plug (CP), inner vessel (IV) and intermediate heat exchanger (IHX) has been carried out using a finite element (FE) code. The approach followed for writing the program (imposing the fall in sodium level (~16 mm/min) at the free surface along with the drop in temperature (~7 K/min) and the creep-fatigue damage assessment for each case has been discussed in this paper.