Analysis of Minor Actinides in Mixed Oxide Fuel Irradiated in Fast Reactor, (I) Determination of Neptunium-237

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著者

    • Otsuka Yuko OTSUKA Yuko
    • Alpha Gamma Section, O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation
    • MOROZUMI Katsufumi
    • Alpha Gamma Section, O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation
    • KONNO Koichi
    • Alpha Gamma Section, O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation
    • KAJITANI Mikio
    • Nuclear Fuel Cycle Development Section, Head Office, Power Reactor and Nuclear Fuel Development Corporation

抄録

The MOX (mixed oxide) fuels, 29.97 wt% (analytical value is 28.08 wt%) plutonium in 8.3 or 12.1 wt% enriched uranium were irradiated up to 120 GWd/t in the MARK-II core of JOYO. The irradiated MOX fuels were dissolved in 8M nitric acid solution. Anion exchange chromatographic procedures were developed for the isolation of neptunium contained in these MOX solutions. Neptunium-237 was separated from the other actinides by using the developed chromatographic technique and the content of <SUP>237</SUP>Np was determined by α spectrometry.<BR>It was found that the content of <SUP>237</SUP>Np slightly increased with increasing burnup. The content did not exceed 0.08 wt% in the 30 wt% MOX fuels irradiated up to about 120 GWd/t. Model calculation by ORIGEN-2 reproduces well the observed content of <SUP>237</SUP>Np.

収録刊行物

  • Journal of nuclear science and technology  

    Journal of nuclear science and technology 35(6), 406-410, 1998-06-25 

    Atomic Energy Society of Japan

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各種コード

  • NII論文ID(NAID)
    10002079525
  • NII書誌ID(NCID)
    AA00703720
  • 本文言語コード
    ENG
  • 資料種別
    ART
  • ISSN
    00223131
  • NDL 記事登録ID
    4503987
  • NDL 雑誌分類
    ZM35(科学技術--物理学)
  • NDL 請求記号
    Z53-A460
  • データ提供元
    CJP書誌  CJP引用  NDL  J-STAGE 
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