• Title/Summary/Keyword: Cask

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Validation of UNIST Monte Carlo code MCS for criticality safety calculations with burnup credit through MOX criticality benchmark problems

  • Ta, Duy Long;Hong, Ser Gi;Lee, Deokjung
    • Nuclear Engineering and Technology
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    • v.53 no.1
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    • pp.19-29
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    • 2021

  • This paper presents the validation of the MCS code for critical safety analysis with burnup credit for the spent fuel casks. The validation process in this work considers five critical benchmark problem sets, which consist of total 80 critical experiments having MOX fuels from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). The similarity analysis with the use of sensitivity and uncertainty tool TSUNAMI in SCALE was used to determine the applicable benchmark experiments corresponding to each spent fuel cask model and then the Upper Safety Limits (USLs) except for the isotopic validation were evaluated following the guidance from NUREG/CR-6698. The validation process in this work was also performed with the MCNP6 for comparison with the results using MCS calculations. The results of this work showed the consistence between MCS and MCNP6 for the MOX fueled criticality benchmarks, thus proving the reliability of the MCS calculations.

  • https://doi.org/10.1016/j.net.2020.06.016 인용 PDF KSCI

Thermal Analysis of a Spent Fuel Storage Cask under Normal and Off-Normal Conditions

  • Lee, J. C.;K. S. Bang;K. S. Seo;Kim, H.D.;Park, B. I.;Lee, H. Y.
    • Proceedings of the Korean Radioactive Waste Society Conference
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    • 2003.11a
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    • pp.601-608
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    • 2003

  • Thermal analyses have been carried out for a spent fuel dry storage cask under normal and off-normal conditions. Environmental temperature is assumed to be $15^{\circ}C$ under the normal condition. The off-normal condition has an environmental temperature of $38^{\circ}C$. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Two of the four air inlet ducts are assumed to be completely blocked. The maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal condition. Temperature distributions for the off-normal conditions were slightly higher than the normal conditions.

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