Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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A Lattice-Based Monte Carlo Evaluation of Canada Deuterium Uranium-6 Safety Parameters

  • Kim, Yonghee (Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST)) ;
  • Hartanto, Donny (Fast Reactor Design Division, Korea Atomic Energy Research Institute (KAERI)) ;
  • Kim, Woosong (Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST))
  • Received : 2015.10.31
  • Accepted : 2016.02.09
  • Published : 2016.06.25
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Abstract

Important safety parameters such as the fuel temperature coefficient (FTC) and the power coefficient of reactivity (PCR) of the CANada Deuterium Uranium (CANDU-6) reactor have been evaluated using the Monte Carlo method. For accurate analysis of the parameters, the Doppler broadening rejection correction scheme was implemented in the MCNPX code to account for the thermal motion of the heavy uranium-238 nucleus in the neutron-U scattering reactions. In this work, a standard fuel lattice has been modeled and the fuel is depleted using MCNPX. The FTC value is evaluated for several burnup points including the mid-burnup representing a near-equilibrium core. The Doppler effect has been evaluated using several cross-section libraries such as ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. The PCR value is also evaluated at mid-burnup conditions to characterize the safety features of an equilibrium CANDU-6 reactor. To improve the reliability of the Monte Carlo calculations, we considered a huge number of neutron histories in this work and the standard deviation of the k-infinity values is only 0.5-1 pcm.

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References

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