Transactions of the Atomic Energy Society of Japan
Online ISSN : 2186-2931
Print ISSN : 1347-2879
ISSN-L : 1347-2879
Application of Dose Evaluation of the MCNP Code for Interim Spent Fuel Cask Storage Facility
Toshiso KOSAKOTakeshi IIMOTOSatoshi ISHIKAWATakafumi TSUBOIMasahiro TERAMURATomomi OKAMURAYoshiyuki NARUMIYA
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2007 Volume 6 Issue 2 Pages 225-238

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Abstract
  The interim storage facility for spent fuel metallic cask is designed as a concrete building structure with air inlet and outlet for circulating the natural cooling. The feature of the interim storage facility is big capacity of spent fuel at several thousands MTU and restricted site usage. It is important to evaluate realistic dose rate in shielding design of the interim storage facility, therefore the three-dimensional continuous-energy Monte Carlo radiation transport code MCNP that exactly treating the complicated geometry was applied. The validation of dose evaluation for interim storage facility by MCNP code were performed by three kinds of neutron shielding benchmark experiments; cask shadow shielding experiment, duct streaming experiment and concrete deep penetration experiment. Dose rate distributions at each benchmark were measured and compared with the calculated results. The comparison showed a good consistency between calculation and experiment results.
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© 2007 Atomic Energy Society of Japan
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