MCNPコードの金属キャスク貯蔵方式中間貯蔵施設線量評価への適用 Application of Dose Evaluation of the MCNP Code for Interim Spent Fuel Cask Storage Facility

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  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.<br>

収録刊行物

  • 日本原子力学会和文論文誌 = Transactions of the Atomic Energy Society of Japan  

    日本原子力学会和文論文誌 = Transactions of the Atomic Energy Society of Japan 6(2), 225-238, 2007-06-25 

    Atomic Energy Society of Japan

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各種コード

  • NII論文ID(NAID)
    10018763751
  • NII書誌ID(NCID)
    AA11643165
  • 本文言語コード
    JPN
  • 資料種別
    ART
  • ISSN
    13472879
  • NDL 記事登録ID
    8884373
  • NDL 雑誌分類
    ZN36(科学技術--原子力工学・工業)
  • NDL 請求記号
    Z74-C788
  • データ提供元
    CJP書誌  NDL  J-STAGE 
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