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.
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Abbreviations
- T Na :
-
Sodium temperature
- T Ar :
-
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
References
K. Natesan et al., Simplified scheme for DHR through steam generators, Internal Report (2015)
R. Sarkar et al., Dependence of control plug life on primary sodium heating rate during power raising, Internal Report (2015)
R. Sarkar et al., Design and life assessment of inner vessel for FBR-1&2. Trans. Indian Inst. Met. 69(2), 543–547 (2016)
R. Srinivasan et al., Conceptual design of intermediate heat exchanger, Internal Report
K. Natesan, Integrated plant operations, Internal Report
RCC-MR Section I, Subsection B, design and construction rules for class-1 components of FBR nuclear islands (2010)
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.
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Sarkar, R., Jaladeen, S., Velusamy, K. (2018). Creep-Fatigue Damage Assessment of Reactor Hot Pool Components During Crash Cooling. In: Prakash, R., Jayaram, V., Saxena, A. (eds) Advances in Structural Integrity. Springer, Singapore. https://doi.org/10.1007/978-981-10-7197-3_28
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DOI: https://doi.org/10.1007/978-981-10-7197-3_28
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