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      KCI등재 SCIE SCOPUS

      Optimization of gap sizes for the high performance of annular nuclear fuels

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      https://www.riss.kr/link?id=A103791055

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      다국어 초록 (Multilingual Abstract)

      Solid-type nuclear fuels have been used for nuclear reactors for a long time. Many countries are currently developing annular fuels toimprove the efficiency of nuclear fuels. The thermoelastic-plastic-creep analyses of solid- and annular-type rods wer...

      Solid-type nuclear fuels have been used for nuclear reactors for a long time. Many countries are currently developing annular fuels toimprove the efficiency of nuclear fuels. The thermoelastic-plastic-creep analyses of solid- and annular-type rods were conducted underthe same conditions. The temperature and stress of the solid- and annular-type rods were compared on the basis of gap size. In this study,we examined the advantages and disadvantages of annular-type fuel regarding the temperature and stress of the pellet and cladding. Theinner and outer gaps between the pellet and cladding play important roles in the temperature and stress distributions of fuel systems.

      Therefore, the optimization of gaps in fuel systems was conducted for a low temperature under certain stress conditions. Thermoelasticplastic-creep analyses were conducted by using an in-house thermoelastic-plastic-creep finite element analysis program in VisualFORTRAN with the effective stress function algorithm. Nonlinear iterative stress analyses were conducted by nonlinear iterative temperatureanalyses; that is, a quasi-fully coupled algorithm was applied to this procedure. In this study, the thermoelastic-plastic-creepanalysis of pressurized water reactor annular fuels was conducted to determine the contacting tendency of the inner–outer gaps betweenthe annular fuel pellets and cladding, as well as to optimize the gap sizes by using the commercial package PIAnO for efficient heat transferat certain stress levels. Most analyses were conducted until the gaps disappeared. However, certain analyses lasted for 1582 days, afterwhich the fuels were replaced.

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      참고문헌 (Reference)

      1 M. Kojic, "The effective stress function algorithm for thermo-elasto-plasticity and creep" 24 (24): 1509-1532, 1987

      2 Y. B. Yang, "Solution strategy and rigid element for nonlinear analysis of elastically structures based on updated Lagrangian formulationk" 29 (29): 1189-1200, 2007

      3 J. M. Winget, "Solution algorithms for nonlinear transient heat conduction analysis employing element-by-element iterative strategies" 52 (52): 711-815, 1985

      4 C. Vitanza, "RIA failure threshold and LOCA limit at high burn-up" 43 (43): 1074-1079, 2006

      5 P. G. Hodge, "Plastic analysis of structures" RE Krieger Publishing Company 1981

      6 "PIAnO User’s Manual Version 3. 5"

      7 N. E. Todreas, "Nuclear systems" CRC Press 2010

      8 M. Suzuki, "Light Water Reactor Fuel Analysis Code FEMAXI-IV(Ver.2)"

      9 M. Ichikawa, "LWR fuel safety research with particular emphasis on RIA/LOCA and other conditions" 26 (26): 118-125, 1989

      10 J. R. Lamarsh, "Introduction to nuclear engineering" Prentice hall 2001

      1 M. Kojic, "The effective stress function algorithm for thermo-elasto-plasticity and creep" 24 (24): 1509-1532, 1987

      2 Y. B. Yang, "Solution strategy and rigid element for nonlinear analysis of elastically structures based on updated Lagrangian formulationk" 29 (29): 1189-1200, 2007

      3 J. M. Winget, "Solution algorithms for nonlinear transient heat conduction analysis employing element-by-element iterative strategies" 52 (52): 711-815, 1985

      4 C. Vitanza, "RIA failure threshold and LOCA limit at high burn-up" 43 (43): 1074-1079, 2006

      5 P. G. Hodge, "Plastic analysis of structures" RE Krieger Publishing Company 1981

      6 "PIAnO User’s Manual Version 3. 5"

      7 N. E. Todreas, "Nuclear systems" CRC Press 2010

      8 M. Suzuki, "Light Water Reactor Fuel Analysis Code FEMAXI-IV(Ver.2)"

      9 M. Ichikawa, "LWR fuel safety research with particular emphasis on RIA/LOCA and other conditions" 26 (26): 118-125, 1989

      10 J. R. Lamarsh, "Introduction to nuclear engineering" Prentice hall 2001

      11 김효찬, "Implementation of Effective-Stress-Function Algorithm for Nuclear Fuel Performance Code" 한국정밀공학회 14 (14): 791-796, 2013

      12 N. Marchal, "Finite element simulation of Pellet-Cladding Interaction (PCI) in nuclear fuel rods" 45 (45): 821-826, 2009

      13 Y. W. Rhee, "Fabrication of sintered annular fuel pellet for HANARO irradiation test" 47 (47): 345-350, 2010

      14 ChangHwan Shin, "Evaluation of loss coefficient for an end plug with side holes in dual-cooled annular nuclear fuel" 대한기계학회 26 (26): 3119-3124, 2012

      15 K. H. Han, "Development of a thermalhydraulic analysis code for annular fuel assemblies" 226 (226): 267-275, 2003

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      학술지 이력

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

      학술지 인용정보
      기준연도 WOS-KCI 통합IF(2년) KCIF(2년) KCIF(3년)
      2016 1.04 0.51 0.84
      KCIF(4년) KCIF(5년) 중심성지수(3년) 즉시성지수
      0.74 0.66 0.369 0.12
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