Temperature Dependence of Creep Properties of Cold-worked Hastelloy XR

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The creep properties of Hastelloy XR, in a solution treated, 10% or 20% cold-worked condition, were investigated at temperatures from 800 to 1, 000°C for the duration of creep tests up to about 2, 500ks. At 800 and 850°C, the steady-state creep rate and rupture ductility decreased and the rupture life increased after cold work of 10% or 20%. Although the rupture life of the 10% cold-worked alloy was longer at 900°C than that of the solution treated one, the rupture lives of the 10% cold-worked and solution treated alloys were almost equal at 950°C, which is the highest helium temperature in an intermediate heat exchanger of the High Temperature Engineering Test Reactor (HTTR). The beneficial effect of 10% cold work on the rupture life and the steady-state creep rate disappeared at 1, 000°C. The beneficial effect of 20% cold work disappeared at 950°C because significant dynamic recrystallization occurred during creep. While rupture ductility of this alloy decreased after cold work of 10% or 20%, it recovered to a considerable extent at 1, 000°C. It is emphasized that these cold work effects should be taken into consideration in design, operation and residual life estimation of high temperature components of the HTTR.

収録刊行物

  • Journal of nuclear science and technology

    Journal of nuclear science and technology 32(6), 539-546, 1995-06-25

    Atomic Energy Society of Japan

参考文献:  14件中 1-14件 を表示

  • <no title>

    Japan Atomic Energy Research Institute

    Present Status of HTGR Research and Development, 1992

    被引用文献2件

  • <no title>

    Japan Atomic Energy Research Institute

    Present Status of HTGR Research and Development, 1993

    被引用文献1件

  • <no title>

    HTTR Des. Lab. Dept. of Fuels and Mater. Res. and Dept. of High-Temperature Eng.

    JAERI-M 89-005, 1989

    被引用文献1件

  • <no title>

    Hada K.

    JAERI-M 90-148, 1990

    被引用文献3件

  • <no title>

    GRANT N. J.

    Trans. ASM 48, 446, 1956

    被引用文献1件

  • <no title>

    LULA R. A.

    Trans. ASME 79, 921, 1957

    被引用文献3件

  • <no title>

    MINO K.

    Tetsu-to-Hagane (J. Iron Steel Inst. Jpn. ) 63, 2372, 1977

    被引用文献1件

  • <no title>

    COOK R. H.

    Nucl. Technol. 66, 283, 1984

    被引用文献1件

  • <no title>

    GAROFALO F.

    Trans. ASM 54, 430, 1961

    被引用文献1件

  • <no title>

    KURATA Y.

    JAERI-M 94-022, 1994

    被引用文献1件

  • <no title>

    SHINDO M.

    Proc. Conf. on Gas-Cooled Reactors Today, Bristol/UK, 1982 (BNES) 2, 179

    被引用文献1件

  • <no title>

    KLUEH R. L.

    J. Nucl. Mater. 79, 363, 1979

    被引用文献1件

  • <no title>

    SCHNEIDER K.

    Nucl. Technol. 66, 289, 1984

    被引用文献1件

  • <no title>

    RICHARDSON G. J.

    Acta Met 14, 1225, 1966

    被引用文献4件

被引用文献:  3件中 1-3件 を表示

各種コード

  • NII論文ID(NAID)
    10002072721
  • 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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