Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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NEUTRONICS INVESTIGATION OF CANADA DEUTERIUM URANIUM 6 REACTOR FUELED (TRANSURANICeTH) O2 USING A COMPUTATIONAL METHOD

  • Received : 2014.08.20
  • Accepted : 2014.11.02
  • Published : 2015.02.25
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Abstract

Background: $^{241}Am$, $^{243}Am$, and $^{237}Np$ isotopes are among the most radiotoxic components of spent nuclear fuel. Recently, researchers have planned different incineration scenarios for the highly radiotoxic elements of nuclear waste in critical reactors. Computational methods are widely used to predict burnup rates of such nuclear wastes that are used under fuel matrixes in critical reactors. Methods: In this work, the Monte Carlo N-particle transport code was used to calculate the neutronic behavior of a transuranic (TRU)-bearing CANada Deuterium Uranium 6 reactor. Results: The computational data showed that the 1.0% TRU-containing thorium-based fuel matrix presents higher proliferation resistance and TRU depletion rate than the other investigated fuel Matrixes. The fuel matrix includes higher negative temperature reactivity coefficients as well. Conclusion: The investigated thorium-based fuel matrix can be successfully used to decrease the production of highly radiotoxic isotopes.

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References

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  2. Coupling of Neutronics and Thermal-Hydraulic Codes for Simulation of the MNSR Reactor vol.193, pp.11, 2019, https://doi.org/10.1080/00295639.2019.1622927