Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Thermal hydraulic analysis of core flow bypass in a typical research reactor

  • Ibrahim, Said M.A. (Mechanical Engineering Department, Faculty of Engineering, Al-Azhar University) ;
  • El-Morshedy, Salah El-Din (Reactors Department, Atomic Energy Authority) ;
  • Abdelmaksoud, Abdelfatah (Reactors Department, Atomic Energy Authority)
  • Received : 2018.04.17
  • Accepted : 2018.08.29
  • Published : 2019.02.25
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Abstract

The main objective of nuclear reactor safety is to maintain the nuclear fuel in a thermally safe condition with enough safety margins during normal operation and anticipated operational occurrences. In this research, core flow bypass is studied under the conditions of the unavailability of safety systems. As core bypass occurs, the core flow rate is assumed to decrease exponentially with a time constant of 25 s to new steady state values of 20, 40, 60, and 80% of the nominal core flow rate. The thermal hydraulic code PARET is used through these calculations. Reactor thermal hydraulic stability is reported for all cases of core flow bypass.

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References

  1. C. Housiadas, Simulation of loss-of-flow transients in research reactors, Ann. Nucl. Energy 27 (18) (2000) 1683-1693. https://doi.org/10.1016/S0306-4549(00)00053-0
  2. F. Muhammad, Simulation of uncontrolled loss of flow transients of a material test research reactor fuelled with low and high enriched uranium dispersion fuels, Ann. Nucl. Energy 37 (4) (2010) 582-591. https://doi.org/10.1016/j.anucene.2009.12.019
  3. H. Kazeminejad, Thermal-hydraulic modeling of flow inversion in a research reactor, Ann. Nucl. Energy 35 (10) (2008) 1813-1819. https://doi.org/10.1016/j.anucene.2008.05.006
  4. S.E.-D. El-Morshedy, Prediction, analysis and solution of the flow inversion phenomenon in a typical MTR-reactor with upward core cooling, Nucl. Eng. Des. 241 (1) (2011) 226-235. https://doi.org/10.1016/j.nucengdes.2010.10.006
  5. S.E.-D. El-Morshedy, Thermal-hydraulic modeling and analysis of a tank in pool reactor for normal operation and loss of flow transient, Prog. Nucl. Energy 61 (2012) 78-87. https://doi.org/10.1016/j.pnucene.2012.07.005
  6. H. Khater, S.E.-D. EL-Morshedy, A. Abdelmaksoud, Simulation of unprotected LOFA in MTR reactors using a mix CFD and one-d computation tool, Ann. Nucl. Energy 83 (2015) 376-385. https://doi.org/10.1016/j.anucene.2015.03.041
  7. C.F. Obenchain, PARET - a Program for the Analysis of Reactor Transients. ACE Research and Development Report, 1969. IDO-1728.
  8. R. Scott Jr., C. Hale, R. Hagen, Transient Tests of Fully Enriched Uranium Oxide Stainless Steel Plate Type C-core in the SPERT-III Reactor. Data Summary Report, 1967. IDO-17223.
  9. W.L. Woodruff, A kinetics and thermal hydraulics capability for the analysis for research reactor, ANL (1983).

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