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

  • Rosy SarkarEmail author
  • S. Jaladeen
  • K. Velusamy
Conference paper


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.


OGDHR Crash cooling CBR Control plug Inner vessel Intermediate heat exchanger 



Sodium temperature


Argon cover gas temperature


Inner surface top length


Inner surface bottom length


Outer surface top length


Outer surface bottom length


Level change


Sodium level



The authors thank Mr. K. Natesan, Mr. Juby Abraham and Mr. U. Parthasarathy of Thermal Hydraulics Section, IGCAR, for providing the temperature input for the reactor components.


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Copyright information

© Springer Nature Singapore Pte Ltd. 2018

Authors and Affiliations

  1. 1.Indira Gandhi Center for Atomic ResearchKalpakkamIndia

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