Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Calculation of fuel temperature profile for heavy water moderated natural uranium oxide fuel using two gas mixture conductance model for noble gas Helium and Xenon

  • Jha, Alok (Homi Bhabha National Institute) ;
  • Gupta, Anurag (HBNI) ;
  • Das, Rajarshi (Station Reactor Physicist, Kakrapar Atomic Power Project NPCIL) ;
  • Paraswar, Shantanu D. (Tarapur Atomic Power Station - 3&4, NPCIL)
  • Received : 2018.11.11
  • Accepted : 2020.05.23
  • Published : 2020.12.25
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Abstract

A model for calculation of fuel temperature profile using binary gas mixture of Helium and Xenon for gap gas conductance is proposed here. In this model, the temperature profile of a fuel pencil from fuel centreline to fuel surface has been calculated by taking into account the dilution of Helium gas filled during fuel manufacturing due to accumulation of fission gas Xenon. In this model an explicit calculation of gap gas conductance of binary gas mixture of Helium and Xenon has been carried out. A computer code Fuel Characteristics Calculator (FCCAL) is developed for the model. The phenomena modelled by FCCAL takes into account heat conduction through the fuel pellet, heat transfer from pellet surface to the cladding through the gap gas and heat transfer from cladding to coolant. The binary noble gas mixture model used in FCCAL is an improvement over the parametric model of Lassmann and Pazdera. The results obtained from the code FCCAL is used for fuel temperature calculation in 3-D neutron diffusion solver for the coolant outlet temperature of the core at steady operation at full power. It is found that there is an improvement in calculation time without compromising accuracy with FCCAL.

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Acknowledgement

The authors(s) wish to acknowledge help of staff at Reactor Physics Section, Tarapur Atomic Power Station for providing access to data for reactor parameters. The authors are thankful to Ranjan Das, Department of E&E IIT Bombay to use Intel Parallel Studio XE ® for FCCAL code testing. The author(s) would also like to thank the Management at Tarapur Atomic Power Station for active encouragement and cooperation in this study.

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