Hydrogen Generation during Cladding/Coolant Interactions under Reactivity Initiated Accident Conditions

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Hydrogen generation in chemical reaction of Zircaloy cladding with coolant water under reactivity initiated accident conditions (RIAs) has been studied with the in-pile experiments in the Nuclear Safety Research Reactor (NSRR). PWR type segmented fuel rods were subjected to pulse irradiation to simulate the power excursion of an RIA. Transient measurements of the void fraction with a densimeter developed for the in-pile experiment detected prompt generation of hydrogen with an increase in cladding surface temperature. The hydrogen generation ceases in the initial 4 seconds of the power excursion even in the experiments resulting in the melting of the fuel cladding and severe damage. The total amount of hydrogen generated during the power burst was estimated by integration of the densimeter data and also by metallographic examinations of the oxidized cladding. The total amount of hydrogen produced increased with an elevation in the maximum temperature of the cladding surface, and the hydrogen generation was intensified as a function of damage to the fuel cladding. The applicability of the PRECIP-II code to the analysis of hydrogen generation under an RIA condition was also examined.

収録刊行物

  • Journal of nuclear science and technology

    Journal of nuclear science and technology 33(1), 43-51, 1996-01-25

    Atomic Energy Society of Japan

参考文献:  17件中 1-17件 を表示

  • <no title>

    Nuclear Safety Commission of Japan

    Safety Evaluation Guidelines 239, 1994

    被引用文献1件

  • <no title>

    IZUTSU S.

    Nucl. Technol. 89, 92, 1990

    被引用文献2件

  • <no title>

    SHIMEGI N.

    J. Nucl. Sci. Technol. 24[3], 203, 1987

    被引用文献1件

  • <no title>

    FUKETA T.

    Nucl. Eng. Des. 146, 181, 1994

    被引用文献1件

  • <no title>

    SHIOZAWA S.

    J. Nucl. Sci. Technol. 19[5], 368, 1982

    被引用文献1件

  • <no title>

    FUJISHIRO T.

    NUREG/CR-0269, TREE-1237, 1978

    被引用文献1件

  • <no title>

    MacDONALD P. E.

    Nucl. Saf. 21[5], 582, 1980

    被引用文献1件

  • <no title>

    SAITO S.

    J. Nucl. Sci. Technol. 14[3], 226, 1977

    被引用文献3件

  • <no title>

    SHIOZAWA S.

    JAERI-M 8187, 1979

    被引用文献1件

  • <no title>

    TSURUTA T.

    J. Nucl. Sci. Technol. 21[7], 515, 1984

    被引用文献1件

  • <no title>

    EPRI NP-195, 1976

    被引用文献1件

  • <no title>

    FUJISHIRO T.

    Proc. Int. Conf. Multiphase Flows '91, Sept. 24-27, Tsukuba, Japan 2, 247, 1991

    被引用文献1件

  • <no title>

    HETSRONI G.

    Handbook of Multiphase Systems, 1982

    被引用文献5件

  • Steam Bubble Behavior During Core Damage Accident of Fast Reactor

    SAKATA K.

    Ph. D thesis, Dept. Nucl. Eng., Univ. of Tokyo, 1982

    被引用文献1件

  • <no title>

    UETSUKA H.

    KfK 3917, 1985

    被引用文献1件

  • <no title>

    SUZUKI M.

    JAERI-M 7720, 1978

    被引用文献1件

  • <no title>

    MALANG S.

    ORNL-5083, 1975

    被引用文献1件

被引用文献:  1件中 1-1件 を表示

各種コード

  • NII論文ID(NAID)
    10002074019
  • NII書誌ID(NCID)
    AA00703720
  • 本文言語コード
    ENG
  • 資料種別
    ART
  • ISSN
    00223131
  • データ提供元
    CJP書誌  CJP引用  J-STAGE 
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