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    Review of the finite element models for a structural integrity evaluation of the sodium-cooled fast reactor high temperature piping

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    https://www.riss.kr/link?id=A103786573

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    다국어 초록 (Multilingual Abstract) kakao i 다국어 번역

    We compared the structural analysis feature of finite element (FE) models for the structural integrity evaluation of the sodium-cooled fast reactor (SFR) high temperature piping and to evaluate the structural integrity against the typical duty cycle event. To evaluate the structural integrity of the high temperature piping per ASME Subsection NH rules, the structural analysis should be carried out first by using a 3-dimensional structural model. The object FE models under consideration in this study are pipe element model, 3-D full model,and 3-D simplified model. The pipe element model is based on the 3-D beam element and effective in understanding overall deformation but less favorable to the detailed stress distribution. The 3-D full model consists of solid structure as well as the contained coolant inside the piping structure with the fluid element. The 3-D simplified model consists of structure shape only, but its material properties are recalculated to reflect the coolant weight effect. The loading conditions for the structural analyses are the mechanical load including dead weight and steady state thermal load. From the analysis results, the piping element model shows the smallest stress intensity, and the required time for FE analysis is also the shortest. The 3-D simplified model shows the most conservative stress intensity output but its calculation time is less than the 3-D full model.
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    We compared the structural analysis feature of finite element (FE) models for the structural integrity evaluation of the sodium-cooled fast reactor (SFR) high temperature piping and to evaluate the structural integrity against the typical duty cycle e...

    We compared the structural analysis feature of finite element (FE) models for the structural integrity evaluation of the sodium-cooled fast reactor (SFR) high temperature piping and to evaluate the structural integrity against the typical duty cycle event. To evaluate the structural integrity of the high temperature piping per ASME Subsection NH rules, the structural analysis should be carried out first by using a 3-dimensional structural model. The object FE models under consideration in this study are pipe element model, 3-D full model,and 3-D simplified model. The pipe element model is based on the 3-D beam element and effective in understanding overall deformation but less favorable to the detailed stress distribution. The 3-D full model consists of solid structure as well as the contained coolant inside the piping structure with the fluid element. The 3-D simplified model consists of structure shape only, but its material properties are recalculated to reflect the coolant weight effect. The loading conditions for the structural analyses are the mechanical load including dead weight and steady state thermal load. From the analysis results, the piping element model shows the smallest stress intensity, and the required time for FE analysis is also the shortest. The 3-D simplified model shows the most conservative stress intensity output but its calculation time is less than the 3-D full model.

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    참고문헌 (Reference)

    1 "PRISM preliminary safety information document, GEFR- 00793"

    2 Dohee Hahn, "KALIMER-600 conceptual design report" 2007

    3 "Fast reactor database 2006 update, IAEA-TECDOC-1531"

    4 Seok-Hoon Kim, "Elevated temperature design evaluations for the ABTR internal structure" 대한기계학회 26 (26): 389-400, 2012

    5 G.-H. Koo, "Development of an ASME-NH program for nuclear component design at elevated temperatures" 85 : 385-393, 2008

    6 C.-G. Park, "Design and structural evaluation of the ABTR IHTS piping for representative duty events of a level a service" 2008

    7 CHANG-GYU PARK, "DESIGN STUDY OF AN IHX SUPPORT STRUCTURE FOR A POOL-TYPE SODIUM-COOLED FAST REACTOR" 한국원자력학회 41 (41): 1323-1332, 2009

    8 구경회, "Creep-fatigue design studies for a sodium-cooled fast reactor with tube sheet-toshell structure subjected to elevated temperature service" 대한기계학회 24 (24): 711-719, 2010

    9 G. H. Koo, "Computer program of SIE ASME-NH code (Revision 1)" Korea Atomic Energy Research Institute 2008

    10 Y. I. Chang, "Advanced burner test reactor preconceptual design report" ANL 2006

    1 "PRISM preliminary safety information document, GEFR- 00793"

    2 Dohee Hahn, "KALIMER-600 conceptual design report" 2007

    3 "Fast reactor database 2006 update, IAEA-TECDOC-1531"

    4 Seok-Hoon Kim, "Elevated temperature design evaluations for the ABTR internal structure" 대한기계학회 26 (26): 389-400, 2012

    5 G.-H. Koo, "Development of an ASME-NH program for nuclear component design at elevated temperatures" 85 : 385-393, 2008

    6 C.-G. Park, "Design and structural evaluation of the ABTR IHTS piping for representative duty events of a level a service" 2008

    7 CHANG-GYU PARK, "DESIGN STUDY OF AN IHX SUPPORT STRUCTURE FOR A POOL-TYPE SODIUM-COOLED FAST REACTOR" 한국원자력학회 41 (41): 1323-1332, 2009

    8 구경회, "Creep-fatigue design studies for a sodium-cooled fast reactor with tube sheet-toshell structure subjected to elevated temperature service" 대한기계학회 24 (24): 711-719, 2010

    9 G. H. Koo, "Computer program of SIE ASME-NH code (Revision 1)" Korea Atomic Energy Research Institute 2008

    10 Y. I. Chang, "Advanced burner test reactor preconceptual design report" ANL 2006

    11 "ASME boiler and pressure vessel code, Section III Rules for Construction of Nuclear Facility Components, Division 1 – Subsection NH, Class 1 Components in Elevated Temperature Service"

    12 "ASME boiler and pressure vessel code, Section III Rules for Construction of Nuclear Facility Components, Division 1 – Subsection NB, Class 1 Components"

    13 "ANSYS user’s manual for revision 11.0"

    14 C.-G. Park, "A comparison study of creep-fatigue defect growth evaluations for a SFR IHTS piping" 2 (2): 20-28, 2008

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    연월일 이력구분 이력상세 등재구분
    2023 평가 해외DB학술지평가 신청대상 (해외등재 학술지 평가)
    2020-01-01 등재 등재학술지 유지 (해외등재 학술지 평가) KCI등재
    2012-11-05 학술지명변경 한글명 : 대한기계학회 영문 논문집 -> Journal of Mechanical Science and Technology KCI등재
    2010-01-01 등재 등재학술지 유지 (등재유지) KCI등재
    2008-01-01 등재 등재학술지 유지 (등재유지) KCI등재
    2006-01-19 학술지명변경 한글명 : KSME International Journal -> 대한기계학회 영문 논문집
    외국어명 : KSME International Journal -> Journal of Mechanical Science and Technology
    KCI등재
    2006-01-01 등재 등재학술지 유지 (등재유지) KCI등재
    2004-01-01 등재 등재학술지 유지 (등재유지) KCI등재
    2001-01-01 등재 등재학술지 선정 (등재후보2차) KCI등재
    1998-07-01 등재 등재후보학술지 선정 (신규평가) KCI등재후보
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