Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Physics Study of Canada Deuterium Uranium Lattice with Coolant Void Reactivity Analysis

  • Park, Jinsu (Ulsan National Institute of Science and Technology) ;
  • Lee, Hyunsuk (Ulsan National Institute of Science and Technology) ;
  • Tak, Taewoo (Ulsan National Institute of Science and Technology) ;
  • Shin, Ho Cheol (Korea Hydro and Nuclear Power Central Research Institute (KHNP-CRI)) ;
  • Lee, Deokjung (Ulsan National Institute of Science and Technology)
  • Received : 2016.02.18
  • Accepted : 2016.07.01
  • Published : 2017.02.25
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Abstract

This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU)-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700) fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the $2{\times}2$ checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

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References

  1. Z. Gholamzadeh, S.M. Mirvakili, H. Khalafi, Neutronics investigation of CANADA Deuterium Uranium 6 reactor fueled (transuranic-Th) $O_2$ using a computational method, Nucl. Eng. Technol 47 (2015) 85-93. https://doi.org/10.1016/j.net.2014.11.002
  2. Atomic Energy of Canada Limited, ACR-700 Technical Description, 10801-01371-TED-001, 2003.
  3. J.H. Bae, J.Y. Jeong, Thermal-hydraulic characteristics for CANFLEX fuel channel using burnable poison in CANDU reactor, Nucl. Eng. Technol 47 (2015) 559-566. https://doi.org/10.1016/j.net.2015.05.001
  4. J.J. Whitlock, W.J. Garland, M.S. Milgram, Effects contributing to positive coolant void reactivity in CANDU, Trans. Am. Nucl. Soc 72 (1995) 329-330.
  5. C.A. Cotton, D. Lee, T.J. Downar, Coolant void reactivity analysis of CANDU and ACR-700 lattices, Trans. Am. Nucl. Soc 90 (2004) 587-589.
  6. C.A. Cotton, D. Lee, T. Kozlowski, T.J. Downar, W.S. Yang, D.E. Carlson, Physics analysis of coolant voiding in the ACR-700 lattice, Trans. Am. Nucl. Soc 92 (2005) 685-687.
  7. MCNP6 User's Manual, LA-CP-13-000634, version 1.0, Los Alamos National Laboratory Report, 2013.
  8. L.L. Carter, R.A. Schwarz, The visual creation and display of MCNP geometries and lattices for criticality problems, Trans. Am. Nucl. Soc 80 (1999). CONF-990605.
  9. J. Choe, H.C. Shin, D. Lee, New burnable absorber for long-cycle low boron operation of PWRs, Ann. Nucl. Energy 88 (2016) 272-279. https://doi.org/10.1016/j.anucene.2015.11.011

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