Analyses of Burnup at plutonium Spots in Uranium-Plutonium Mixed Oxide Fuels in Light Water Reactors by Neutron Transport and Burnup Calculations

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Plutonium concentrations and burnup at Pu spots were calculated in U-Pu mixed oxide (MOX) fuel pellets for light water reactors with the neutron transport and burnup calculation code VIMBURN. The calculation models were suggested for Pu spots and U matrices in a heterogeneous MOX fuel pellet. The calculated Pu concentrations and burnup at Pu spots were compared with the PIEs data in a MOX pellet (38.8MWd/kgHM). The calculated Pu concentrations agreed by 5-18% with the measured ones, and the calculated burnup did by less than 10% with the estimated one with the measured Nd concentrations. Commercial PWR types of MOX fuels were also analyzed with the calculation code and the models. Burnup at Pu spot increased as the distance was greater from the radial center of a MOX fuel pellet. Burnup at Pu spots in the peripheral region became 3-5 times higher than pellet average burnup of 40MWd/kgHM. The diameters (20-100μm) of Pu spots were not found a significant factor for burnup at Pu spots. In the outer half volume region (outer than r/r<SUB>0</SUB>=0.7) of a MOX fuel pellet, burnup at Pu spots exceeded 70MWd/kgHM (the threshold burnup of microstructure change in UO<SUB>2</SUB> fuel pellet) at pellet average burnup of 14-30MWd/kgHM.

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  • Journal of nuclear science and technology  

    Journal of nuclear science and technology 34(6), 551-558, 1997-06-25 

    Atomic Energy Society of Japan

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各種コード

  • NII論文ID(NAID)
    10002077319
  • 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
    4240102
  • 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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