Nuclear Engineering and Technology

  • 제49권4호
  • /
  • Pages.700-709
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  • 2017
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  • 1738-5733(pISSN)
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  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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Effect of Kinetic Parameters on Simultaneous Ramp Reactivity Insertion Plus Beam Tube Flooding Accident in a Typical Low Enriched U3Si2-Al Fuel-Based Material Testing Reactor-Type Research Reactor

  • Nasir, Rubina (Department of Physics, Air University) ;
  • Mirza, Sikander M. (Department of Physics & Applied Mathematics, Pakistan Institute of Engineering & Applied Sciences) ;
  • Mirza, Nasir M. (Department of Physics & Applied Mathematics, Pakistan Institute of Engineering & Applied Sciences)
  • 투고 : 2016.08.22
  • 심사 : 2016.12.20
  • 발행 : 2017.08.25
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초록

This work looks at the effect of changes in kinetic parameters on simultaneous reactivity insertions and beam tube flooding in a typical material testing reactor-type research reactor with low enriched high density ($U_3Si_2-Al$) fuel. Using a modified PARET code, various ramp reactivity insertions (from $0.1/0.5 s to $1.3/0.5 s) plus beam tube flooding ($0.5/0.25 s) accidents under uncontrolled conditions were analyzed to find their effects on peak power, net reactivity, and temperature. Then, the effects of changes in kinetic parameters including the Doppler coefficient, prompt neutron lifetime, and delayed neutron fractions on simultaneous reactivity insertion and beam tube flooding accidents were analyzed. Results show that the power peak values are significantly sensitive to the Doppler coefficient of the system in coupled accidents. The material testing reactor-type system under such a coupled accident is not very sensitive to changes in the prompt neutron life time; the core under such a coupled transient is not very sensitive to changes in the effective delayed neutron fraction.

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참고문헌

  1. R.K. Cardell, D.H. Herborn, J.E. Houghtaling, Reactivity accident test results and analysis for the SPERT IIIE Core, IDO-17281, 1967.
  2. D.E. Cullen, R. Muranaka, J. Schmidt, Applications in Nuclear Data and Reactor Physics, Word Scientific, Vienna, Austria, 1986.
  3. C.F. Obenchain, PARET-A Program for the Analysis of Reactor Transients, AEC Research and Development Report, IDO-17282, USAEC, Idaho, 1969.
  4. W.L. Woodruff, The PARET Code and the Analysis of the SPERT-I Transients, ANL/RERTR/TM-4, 1982.
  5. C. Housiadas, Simulation of loss of flow transients in research reactors, Ann. Nucl. Energy 27 (2000) 1683-1693. https://doi.org/10.1016/S0306-4549(00)00053-0
  6. M. Iqbal, N.M. Mirza, S.M. Mirza, S.K. Ayyazuddin, Study of the void coefficient of reactivity in a typical pool-type research reactor, Ann. Nucl. Energy 24 (1997) 177-186. https://doi.org/10.1016/0306-4549(96)00007-2
  7. J.E. Metos, E.M. Pennington, K.E. Freese, W.L. Woodruff, Safety-related benchmark calculations for MTR type reactors with HEU, MEU and LEU fuels, Research Reactor Core Conversion Guidebook, Volume 3, IAEA, Vienna, 1992.
  8. A.M. Mirza, S. Khanam, N.M. Mirza, Simulation of reactivity transients in current MTRs, Ann. Nucl. Energy 25 (1998) 1465-1484. https://doi.org/10.1016/S0306-4549(98)00020-6
  9. S.M. Mirza, Simulation of over-power transients in tank-in-pool type research reactors, Ann. Nucl. Energy 24 (1997) 871-881. https://doi.org/10.1016/S0306-4549(96)00079-5
  10. R. Nasir, N.M. Mirza, S.M. Mirza, Sensitivity of reactivity insertion limits with respect to safety parameters in a typical MTR, Ann. Nucl. Energy 26 (1999) 1517-1535. https://doi.org/10.1016/S0306-4549(99)00038-9
  11. A. Salama, CFD analysis of fast loss-of-flow accident in typical MTR reactor undergoing partial and full blockage: the average channel scenario, Prog. Nucl. Energ. 60 (2012) 1-13. https://doi.org/10.1016/j.pnucene.2012.05.002
  12. M.A. Gaheen, S. Elaraby, M.N. Aly, M.S. Nagy, Simulation and analysis of IAEA benchmark transients, Prog. Nucl. Energ. 49 (2007) 217-229. https://doi.org/10.1016/j.pnucene.2006.12.003
  13. H. Khater, T. Abu-El-Maty, E.S. El-Morshdy, Thermal-hydraulic modeling of reactivity accident in MTR reactors, Ann. Nucl. Energy 34 (2007) 732-742. https://doi.org/10.1016/j.anucene.2007.03.012
  14. A. Hainoun, N. Ghazi, B.M. Abdu-Moaiz, Safety analysis of the reference research reactor MTR during reactivity insertion accident using the code MERSAT, Ann. Nucl. Energy 37 (2010) 853-860. https://doi.org/10.1016/j.anucene.2010.02.013
  15. T. Hamidouche, A. Bousbia-Salah, Assessment of RELAP5 point kinetic model against reactivity insertion transient in the IAEA 10 MW MTR research reactor, Nucl. Eng. Des. 240 (2010) 672-677. https://doi.org/10.1016/j.nucengdes.2009.11.002
  16. R. Nasir, M.K. Butt, S.M. Mirza, N.M. Mirza, Effect of high density dispersion fuels on transient behavior of MTR type research reactor under multiple reactivity transients, Prog. Nucl. Energ. 85 (2015) 511-517. https://doi.org/10.1016/j.pnucene.2015.07.018
  17. R. Nasir, M.K. Butt, S.M. Mirza, N.M. Mirza, Simultaneous multiple reactivity insertions in a typical MTR-type research reactor having U3Si2-Al fuel, Ann. Nucl. Energy 85 (2015) 869-878. https://doi.org/10.1016/j.anucene.2015.07.003
  18. IAEA, Research Reactor Core Conversion Guidebook, Volume-3: IAEA TECDOC-643, International Atomic Energy Agency, Vienna, 1992.
  19. X-5 Monte Carlo Team, MCNP-A General Monte Carlo NParticle Transport Code, Version 5, User's Guide, 2003.
  20. A. Khakim, Safety analysis of MTR type research reactor during postulated beam tube break including positive reactivity, Ann. Nucl. Energy 87 (2016) 793-798. https://doi.org/10.1016/j.anucene.2014.08.018
  21. W.L. Woodruff, N.A. Hanan, R.S. Smith, J.E. Matos, A Comparison of the PARET/ANL and RELAP5/MOD3 Codes for the Analysis of IAEA Benchmark Transients, Proc.. of 1996 Int. Mtg. On Reduced Enrichment for Research and Test Reactors, Seoul, Korea, 1996, pp. 260-269.
  22. S. Bakhtyar, A. Salahuddin, Neutronic analysis of PARR-1 equilibrium core, PINSTECH-164, RPG/NED, Pakistan Institute of Nuclear Science & Technology, 1999.
  23. W.L. Woodruff, Evaluation and selection of hot channel (peaking) factors for research reactor applications, ANL/RERTR/TM-28, Argonne National Laboratory, Illinois, 1997.
  24. E.E. Lewis, Nuclear Power Reactor Safety, John Wiley, New York, 1979.
  25. K.O. Ott, R.J. Neuhold, Introductory Nuclear Reactor Dynamics, American Nuclear Society, Illinois, 1985.
  26. R. Nasir, N.M. Mirza, S.M. Mirza, Study of successive ramp reactivity insertions in typical pool-type research reactors, Prog. Nucl. Energ. 66 (2013) 115-123. https://doi.org/10.1016/j.pnucene.2013.03.021
  27. F. Muhammad, Kinetic parameters of a low enriched uranium fuelled material test research reactor at end-of-life, Ann. Nucl. Energy 37 (2010) 1411-1414. https://doi.org/10.1016/j.anucene.2010.05.012
  28. G.L. Hofman [Internet]. A short note of high density dispersion fuel, Argonne National Laboratory, 1996 [cited 2008 Jun 28]. Available from: www.rertr.anl.gov/ADVFUELS/GHHD.PDF.
  29. M.A. Pope, R. Sonat Sen, A.M. Ougouag, G. Youinou, B. Boer, Neutronic analysis of the burning of transuranics in fully ceramic micro-encapsulated tri-isotropic particle-fuel in a PWR, Nucl. Eng. Des. 252 (2012) 215-225. https://doi.org/10.1016/j.nucengdes.2012.07.013

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