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      KCI등재 SCIE SCOPUS

      Coupled irradiation-thermal-mechanical analysis of the solid-state core in a heat pipe cooled reactor

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      https://www.riss.kr/link?id=A108146114

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      다국어 초록 (Multilingual Abstract)

      The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K withthermal and irradiation-induced expansion during burnup. The expansion changes the gap thicknessbetween the solid components and the material properties, ...

      The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K withthermal and irradiation-induced expansion during burnup. The expansion changes the gap thicknessbetween the solid components and the material properties, and may even cause the gap closure, whichthen significantly influences the thermal and mechanical characteristics of the reactor core. This studydeveloped an irradiation behavior model for HPRTRAN, a heat pipe reactor system analysis code, tointroduce the irradiation effects such as swelling and creep. The megawatt heat pipe reactor MegaPowerwas chosen as an application case. The coupled irradiation-thermal-mechanical model was developed tosimulate the irradiation effects on the heat transfer and stresses of the whole reactor core. The resultsshow that the irradiation deformation effect is significant, with the irradiation-induced strains up to2.82% for fuel and 0.30% for monolith at the end of the reactor lifetime. The peak temperatures during thelifetime are 1027:3 K for the fuel and 956:2 K for monolith. The gap closure enhances the heat transferbut caused high stresses exceeding the yield strength in the monolith

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      참고문헌 (Reference)

      1 Y. Ma, "Transient heat pipe failure accident analysis of a megawatt heat pipe cooled reactor" 140 : 103904-, 2021

      2 M. R. Eslami, "Theory of Elasticity and Thermal Stresses" Springer 2013

      3 B. H. Yan, "The technology of micro heat pipe cooled reactor : a review" 135 : 106948-, 2020

      4 H. Cao, "The research on the heat transfer of a solid-core nuclear reactor cooled by heat pipe through a numerical simulation, considering the assembly gaps" 130 : 431-439, 2019

      5 D. I. Poston, "The Heatpipe-Operated Mars Exploration Reactor (HOMER)" AIP, American Institute of Physics 2001

      6 P. Hu, "Steady-state thermal analysis of a fuel module in the heat pipe-cooled reactor(in Chinese)" 4 : 375-378, 2013

      7 J.W. Sterbentz, "Special purpose nuclear reactor (5 MW) for reliable power at remote sites assessment report" Idaho National Lab (INL) 2017

      8 B. Mihaila, "Simulations of coupled heat transport, oxygen diffusion, and thermal expansion in UO2nuclear fuel elements" 394 : 182-189, 2009

      9 Y. Zhao, "Simulation of the irradiation-induced thermo-mechanical behaviors evolution in monolithic UeMo/Zr fuel plates under a heterogeneous irradiation condition" 285 : 84-97, 2015

      10 K. Wang, "RMC e a Monte Carlo code for reactor core analysis" 82 : 121-129, 2015

      1 Y. Ma, "Transient heat pipe failure accident analysis of a megawatt heat pipe cooled reactor" 140 : 103904-, 2021

      2 M. R. Eslami, "Theory of Elasticity and Thermal Stresses" Springer 2013

      3 B. H. Yan, "The technology of micro heat pipe cooled reactor : a review" 135 : 106948-, 2020

      4 H. Cao, "The research on the heat transfer of a solid-core nuclear reactor cooled by heat pipe through a numerical simulation, considering the assembly gaps" 130 : 431-439, 2019

      5 D. I. Poston, "The Heatpipe-Operated Mars Exploration Reactor (HOMER)" AIP, American Institute of Physics 2001

      6 P. Hu, "Steady-state thermal analysis of a fuel module in the heat pipe-cooled reactor(in Chinese)" 4 : 375-378, 2013

      7 J.W. Sterbentz, "Special purpose nuclear reactor (5 MW) for reliable power at remote sites assessment report" Idaho National Lab (INL) 2017

      8 B. Mihaila, "Simulations of coupled heat transport, oxygen diffusion, and thermal expansion in UO2nuclear fuel elements" 394 : 182-189, 2009

      9 Y. Zhao, "Simulation of the irradiation-induced thermo-mechanical behaviors evolution in monolithic UeMo/Zr fuel plates under a heterogeneous irradiation condition" 285 : 84-97, 2015

      10 K. Wang, "RMC e a Monte Carlo code for reactor core analysis" 82 : 121-129, 2015

      11 C. Tang, "Preliminary research on the irradiationthermal-mechanical coupling behavior simulation method of FCM fuel" 1 : 51-56, 2019

      12 Y. Ma, "Neutronic and thermal-mechanical coupling analyses in a solid-state reactor using Monte Carlo and finite element methods" 151 : 107923-, 2021

      13 W.G. Luscher, "Material property correlations: comparisons between FRAPCON-3.4, FRAPTRAN 1.4, and MATPRO" Pacific Northwest National Lab 2010

      14 D.L. Hagrman, "MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior" Idaho National Engineering Lab 1979

      15 M.K. Meyer ; J.GAN ; D.D. KEISER ; E. PEREZ ; A. ROBINSON ; D.M. WACHS ; N. WOOLSTENHULME ; G.L. Hofman ; Y. S. Kim, "Irradiation Performance of U-Mo Monolithic Fuel" 한국원자력학회 46 (46): 169-182, 2014

      16 H. Yu, "Initiation and development of heat pipe cooled reactor" 40 : 1-8, 2019

      17 P. Li, "Inhomogeneous interface structure and mechanical properties of rotary friction welded TC4 titanium alloy/316L stainless steel joints" 33 : 54-63, 2018

      18 Y. Ma, "Heat pipe failure accident analysis in megawatt heat pipe cooled reactor" 149 : 107755-, 2020

      19 A. Ross, "Heat Transfer Coefficient between UO 2 and Zircaloy-2" Atomic Energy of Canada Limited 1962

      20 K. Une, "Fuel oxidation and irradiation behaviors of defective BWR fuel rods" 223 : 40-50, 1995

      21 Y. Rashid, "Fuel Analysis and Licensing Code : FALCON MOD01" EPRI 2004

      22 Y. S. Kim, "Fission product induced swelling of UeMo alloy fuel" 419 : 291-301, 2011

      23 J. Rest, "Fission gas release from UO2 nuclear fuel : a review" 513 : 310-345, 2019

      24 K. C. Mills, "Equations for the calculation of the thermophysical properties of stainless steel" 44 : 1661-1668, 2004

      25 H. Yu, "Development of fuel rod behavior analysis code(FROBA)and its application to AP1000" 50 : 8-17, 2012

      26 P.R. Mcclure, "Design of megawatt power level heat pipe reactors" Los Alamos National Lab (LANL) 2015

      27 P. McClure, "Design of megawatt power level heat pipe reactors" Los Alamos National Lab 2015

      28 P. Van Uffelen, "Comprehensive Nuclear Materials" Elsevier 535-577, 2012

      29 L. Bernard, "An efficient model for the analysis of fission gas release" 302 : 125-134, 2002

      30 Y. Ma, "A transient and steady-state network model forannular-wick heat pipes in continuum flow pattern" 2019

      31 P. Lucuta, "A pragmatic approach to modelling thermal conductivity of irradiated UO2 fuel : review and recommendations" 232 : 166-180, 1996

      32 Z. J. Zuo, "A network thermodynamic analysis of the heat pipe" 41 : 1473-1484, 1998

      33 A. Booth, "A method of calculating fission gas diffusion from UO 2 fuel and its application to the X-2-f loop" Atomic Energy of Canada Limited 1957

      34 K. Geelhood, "A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup" Pacific Northwest, National Laboratory 2015

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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
      2014-01-01 평가 SCIE 등재 (등재유지) KCI등재
      2014-01-01 평가 SCOPUS 등재 (등재유지) KCI등재
      2011-01-01 평가 등재학술지 유지 (등재유지) KCI등재
      2009-01-01 평가 등재학술지 유지 (등재유지) KCI등재
      2007-01-01 평가 등재학술지 유지 (등재유지) KCI등재
      2006-07-31 학술지명변경 한글명 : Jorunal of the Korean Nuclear Society -> Nuclear Engineering and Technology
      외국어명 : 미등록 -> Nuclear Engineering and Technology
      KCI등재후보
      2004-01-01 평가 등재후보학술지 선정 (신규평가) KCI등재후보
      2003-01-01 평가 등재후보 1차 PASS (등재후보1차) KCI등재후보
      2002-01-01 평가 등재후보학술지 유지 (등재후보1차) KCI등재후보
      1999-01-01 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 0.17 0.77
      KCIF(4년) KCIF(5년) 중심성지수(3년) 즉시성지수
      0.63 0.56 0.343 0.11
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