Intentional Depressurization of Steam Generator Secondary Side during a PWR Small-Break Loss-Coolant Accident

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The consequence of intentional depressurization of the steam generator (SG) secondary side during a small break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR) was studied using the ROSA-IV Large Scale Test Facility (LSTF). The LSTF is a 1:48 volumetrically scaled full-height model of a Westinghouse-type PWR. The experiment simu- lated a 0.5% cold leg small-break LOCA with total failure of high pressure injection (HPI) system. The SG secondary-side atmospheric relief valve (ARV) was latched open after the core top region was uncovered. Then, the primary pressure closely followed the secondary pressure and dropped to the accumulator (ACC) injection pressure of 4.51MPa before the core became severely overheated. A post test analysis, which was performed using the RELAP5/ MOD2 code with modifications made by the authors, predicted well the primary and secondary side responses during this experiment.

収録刊行物

  • Journal of nuclear science and technology

    Journal of nuclear science and technology 32(2), 101-110, 1995-02-25

    Atomic Energy Society of Japan

参考文献:  12件中 1-12件 を表示

  • <no title>

    LOOMIS G. G.

    NUREG/CR-4945, EGG-2509, 1987

    被引用文献1件

  • <no title>

    CLEMENT P.

    OECD/NEA/CSNI/R(92)20, 1992

    被引用文献1件

  • <no title>

    ROSA-VI Group

    JAERI-M 84-237, 1985

    被引用文献1件

  • <no title>

    DIMENNA R. A.

    NURGE/CR-5194, EGG-2531, 1988

    被引用文献1件

  • <no title>

    FAUSKE H. K.

    Heat Transfer-Cleveland, AIChE Symp. Ser. 59, (61), 1965

    被引用文献1件

  • Heat Transfer-Pittsburgh 1987

    KUKITA Y.

    Heat Transfer Cleveland AIChE Symp. Ser. 257, (83), 212-217, 1987

    被引用文献1件

  • <no title>

    ASAKA H.

    Proc. 1st Int. Conf. on Supercomputing in Nuclear Applications, Mito, Japan, Mar. 12-16, 210-217, 1990

    被引用文献1件

  • <no title>

    YONOMOTO T.

    Proc. ANS Int. Topical Mtg. on Safety of Thermal Reactors, Portland, Oregon, USA, July 21-25, 522-529, 1991

    被引用文献1件

  • <no title>

    ARDRON K. H.

    Nucl. Eng. Des. 39, 257-266, 1976

    被引用文献3件

  • <no title>

    YONOMOTO T.

    J. Nucl. Sci. Technol. 25[5], 441-455, 1988

    被引用文献1件

  • <no title>

    ASAKA H.

    Exp. Therm, Fluid Sci. 3, 588-596, 1990

    被引用文献3件

  • <no title>

    SALLET D. W.

    Nucl. Sci. Eng. 88, 220-244, 1984

    被引用文献1件

被引用文献:  2件中 1-2件 を表示

各種コード

  • NII論文ID(NAID)
    10002072016
  • 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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