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

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

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.

Keywords

OGDHR Crash cooling CBR Control plug Inner vessel Intermediate heat exchanger 

Nomenclature

TNa

Sodium temperature

TAr

Argon cover gas temperature

lti

Inner surface top length

lbi

Inner surface bottom length

lto

Outer surface top length

lbo

Outer surface bottom length

L

Level change

L

Sodium level

Notes

Acknowledgements

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.

References

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    K. Natesan et al., Simplified scheme for DHR through steam generators, Internal Report (2015)Google Scholar
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    R. Sarkar et al., Dependence of control plug life on primary sodium heating rate during power raising, Internal Report (2015)Google Scholar
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    R. Sarkar et al., Design and life assessment of inner vessel for FBR-1&2. Trans. Indian Inst. Met. 69(2), 543–547 (2016)CrossRefGoogle Scholar
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    R. Srinivasan et al., Conceptual design of intermediate heat exchanger, Internal ReportGoogle Scholar
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    K. Natesan, Integrated plant operations, Internal ReportGoogle Scholar
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    RCC-MR Section I, Subsection B, design and construction rules for class-1 components of FBR nuclear islands (2010)Google Scholar

Copyright information

© Springer Nature Singapore Pte Ltd. 2018

Authors and Affiliations

  1. 1.Indira Gandhi Center for Atomic ResearchKalpakkamIndia

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