Nuclear Engineering and Technology

  • 제52권9호
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  • Pages.1907-1916
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  • 2020
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  • 1738-5733(pISSN)
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  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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Validation of nuclide depletion capabilities in Monte Carlo code MCS

  • Ebiwonjumi, Bamidele (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology) ;
  • Lee, Hyunsuk (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology) ;
  • Kim, Wonkyeong (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology) ;
  • Lee, Deokjung (Department of Nuclear Engineering, Ulsan National Institute of Science and Technology)
  • 투고 : 2019.03.14
  • 심사 : 2020.02.25
  • 발행 : 2020.09.25
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초록

In this work, the depletion capability implemented in Monte Carlo code MCS is investigated to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The depletion analysis is performed with the ENDF/B-VII.1 library and a fuel assembly model. The transmutation equation is solved by the Chebyshev Rational Approximation Method (CRAM) with a depletion chain of 3820 isotopes. 18 actinides and 19 fission products are analyzed in 14 SNF samples. The effect of statistical uncertainties on the calculated number densities is discussed. On average, most of the actinides and fission products analyzed are predicted within ±6% of the experiment. MCS depletion results are also compared to other depletion codes based on publicly reported information in literature. The code-to-code analysis shows comparable accuracy. Overall, it is demonstrated that the depletion capability in MCS can be reliably applied in the prediction of SNF isotopic inventory.

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참고문헌

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피인용 문헌

  1. Assessment of the Radiotoxicity of Spent Nuclear Fuel from a Fleet of PWR Reactors vol.14, pp.11, 2020, https://doi.org/10.3390/en14113094