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    Mechanical Properties of Roll Extruded Nuclear Reactor Piping

    Source: Journal of Pressure Vessel Technology:;1976:;volume( 098 ):;issue: 002::page 105
    Author:
    J. M. Steichen
    ,
    R. L. Knecht
    DOI: 10.1115/1.3454346
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: The elevated temperature mechanical properties of large diameter (28 in.) seamless pipe produced by roll extrusion for use as primary piping for sodium coolant in the Fast Flux Test Facility (FFTF) have been characterized. The three heats of type 316H stainless steel piping material used in this study exhibited very consistent mechanical properties and chemical compositions. Tensile and creep-rupture properties exceed values on which the allowable stresses for ASME Code Case 1592 on Nuclear Components in Elevated Temperature Service were based. Tensile strength and ductility were essentially unchanged by aging in static sodium at 1050°F (566°C) for times to 10,000 hr. High strain rate tensile tests showed that tensile properties were insensitive to strain rate at temperatures to 900°F (482°C) and that for temperatures of 1050°F (566°C) and above both strength and ductility significantly increased with increasing strain rate. Fatigue-crack propagation properties were comparable to results obtained on plate material and no differences in crack propagation were found between axial and circumferential orientations.
    keyword(s): Mechanical properties , Pipes , Nuclear reactors , Temperature , Ductility , Sodium , Stainless steel , Tensile strength , Test facilities , ASME Standards , Creep , Rupture , Extruding , Stress , Coolants , Crack propagation AND Fatigue cracks ,
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      Mechanical Properties of Roll Extruded Nuclear Reactor Piping

    URI
    http://yetl.yabesh.ir/yetl1/handle/yetl/89198
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    contributor authorJ. M. Steichen
    contributor authorR. L. Knecht
    date accessioned2017-05-08T23:01:41Z
    date available2017-05-08T23:01:41Z
    date copyrightMay, 1976
    date issued1976
    identifier issn0094-9930
    identifier otherJPVTAS-28131#105_1.pdf
    identifier urihttp://yetl.yabesh.ir/yetl/handle/yetl/89198
    description abstractThe elevated temperature mechanical properties of large diameter (28 in.) seamless pipe produced by roll extrusion for use as primary piping for sodium coolant in the Fast Flux Test Facility (FFTF) have been characterized. The three heats of type 316H stainless steel piping material used in this study exhibited very consistent mechanical properties and chemical compositions. Tensile and creep-rupture properties exceed values on which the allowable stresses for ASME Code Case 1592 on Nuclear Components in Elevated Temperature Service were based. Tensile strength and ductility were essentially unchanged by aging in static sodium at 1050°F (566°C) for times to 10,000 hr. High strain rate tensile tests showed that tensile properties were insensitive to strain rate at temperatures to 900°F (482°C) and that for temperatures of 1050°F (566°C) and above both strength and ductility significantly increased with increasing strain rate. Fatigue-crack propagation properties were comparable to results obtained on plate material and no differences in crack propagation were found between axial and circumferential orientations.
    publisherThe American Society of Mechanical Engineers (ASME)
    titleMechanical Properties of Roll Extruded Nuclear Reactor Piping
    typeJournal Paper
    journal volume98
    journal issue2
    journal titleJournal of Pressure Vessel Technology
    identifier doi10.1115/1.3454346
    journal fristpage105
    journal lastpage110
    identifier eissn1528-8978
    keywordsMechanical properties
    keywordsPipes
    keywordsNuclear reactors
    keywordsTemperature
    keywordsDuctility
    keywordsSodium
    keywordsStainless steel
    keywordsTensile strength
    keywordsTest facilities
    keywordsASME Standards
    keywordsCreep
    keywordsRupture
    keywordsExtruding
    keywordsStress
    keywordsCoolants
    keywordsCrack propagation AND Fatigue cracks
    treeJournal of Pressure Vessel Technology:;1976:;volume( 098 ):;issue: 002
    contenttypeFulltext
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