球床反应堆燃料循环球流管系结构过球可靠性分析
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摘要
基于结构概率极限状态设计理论提出了一种适用于球床反应堆燃料循环球流管系结构过球可靠性分析的方法。建立了管道过球功能函数,以及过球可靠性分析的量化指标——管道结构过球可靠性指标β。利用有限元方法建立管系整体梁单元模型,并进行管系静力及动力分析,结合应力分析结果计算β,并通过目标指标β*控制球流管道变形。工程设计分析实例表明:所提出的球流管系过球可靠性分析方法能够在满足管系承载安全性要求的同时提高整体过球可靠性。
A method was developed to analyze the fuel flow reliability of fuel spheres in the recycling pipeline of a pebble bed reactor based on probability limit design theory.The fuel flow and its reliability index,β,were derived to quantify the fuel-passing reliability analysis.A finite element model of the pipeline was used to analyze the static and dynamic processes.The reliability index,β,was calculated using the results of stress calculations.The deformation of the flow passage in the pipeline was controlled by a target index β*.Application of the method to an engineering design shows that this method ensures both the bearing security and the fuel flow reliability of a fuel recycling pipeline in a pebble bed reactor.
A method was developed to analyze the fuel flow reliability of fuel spheres in the recycling pipeline of a pebble bed reactor based on probability limit design theory.The fuel flow and its reliability index,β,were derived to quantify the fuel-passing reliability analysis.A finite element model of the pipeline was used to analyze the static and dynamic processes.The reliability index,β,was calculated using the results of stress calculations.The deformation of the flow passage in the pipeline was controlled by a target index β*.Application of the method to an engineering design shows that this method ensures both the bearing security and the fuel flow reliability of a fuel recycling pipeline in a pebble bed reactor.
引文
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[9]戴守通,汪军,韩治,等.中国先进研究堆主回路进堆管系接管载荷优化[J].原子能科学技术,2009,42(S2):490-494.DAI Shoutong,WANG Jun,HAN Zhi,et al.Nozzle loadsoptimizing of inlet primary loop piping in China advancedresearch reactor[J].Atomic Energy Science andTechnology,2009,42(S2):490-494.(in Chinese)
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[12]Slagis G.Recommended revisions to nuclear piping thermalexpansion stress limits[J].Journal of Pressure VesselTechnology,Transactions of the ASME,2010,133(2):021102-1-7.
[2]刘继国,孙德刚,关廷光,等.10MW高温气冷实验堆燃料元件装卸系统设计[J].高技术通讯.1996,6(5):51-55.LIU Jiguo,SUN Degang,GUAN Tingguang,et al.Designof 10 MW HTR-TM fuel handling system[J].HighTechnology Letters,1996,6(5):51-55.(in Chinese)
[3]Avrithi K,Ayyub B M.Load and resistance factor design(LRFD)of nuclear straight pipes for loads that cause primarystress[J].Journal of Pressure Vessel Technology,Transactions of the ASME,2010,132(2):021101-1-14.
[4]GB50153-2008.工程结构可靠性设计统一标准[S].北京:中国建筑工业出版社,2008.GB 50153-2008.Unified Standard for Reliability Design ofEngineering Structures[S].Beijing:China Architecture&Building Press,2008.(in Chinese)
[5]张征明,王敏稚,何树延.HTR-10核安全一级管道的力学分析[J].清华大学学报:自然科学版,2000,40(12):14-17.ZHANG Zhengming,WANG Minzhi,HE Shuyan.Mechanical analysis of the nuclear class 1piping in HTR-10[J].J Tsinghua Univ:Sci&Tech,2000,40(12):14-17.(in Chinese)
[6]刘爱文,张素灵,胡聿贤,等.地震断层作用下埋地管线的反应分析[J].地震工程与工程振动,2002,22(2):22-27.LIU Aiwen,ZHANG Suling,HU Yuxian,et al.A methodfor analyzing response of buried pipeline due to earthquakefault movement[J].Earthquake Engineering andEngineering Vibration,2002,22(2):22-27.(in Chinese)
[7]董建令,张晓航,殷德健,等.10MW高温气冷堆蒸汽发生器传热管束应力分析[J].核动力工程,2001,22(5):433-437.DONG Jianling,ZHANG Xiaohang,YIN Dejian,et al.Stress analysis of HTR-10steam generator heat exchangingtubes[J].Nuclear Power Engineering,2000,22(5):433-437.(in Chinese)
[8]张丽,何树延,吴莘馨,等.HTR-PM热气导管结构强度的分析与计算[J].原子能科学技术,2009,43(S2):271-274.ZHANG Li,HE Shuyan,WU Xinxin,et al.Analysis andcalculation of structure strength of hot gas duct in HTR-PM[J].Atomic Energy Science and Technology,2009,43(S2):271-274.(in Chinese)
[9]戴守通,汪军,韩治,等.中国先进研究堆主回路进堆管系接管载荷优化[J].原子能科学技术,2009,42(S2):490-494.DAI Shoutong,WANG Jun,HAN Zhi,et al.Nozzle loadsoptimizing of inlet primary loop piping in China advancedresearch reactor[J].Atomic Energy Science andTechnology,2009,42(S2):490-494.(in Chinese)
[10]Working Group on Piping of the Design Subgroup of theNuclear Power Committee of the ASME Boiler and PressureVessel Committee.Sample Analysis of a Piping System:Class 1Nuclear[R].New York:ASME,2004.
[11]ASME-BPVC-Ⅲ-1-NC.Rules for Construction of NuclearPower Plant Components,Division 1:Subsection NC[S].New York:ASME,2004.
[12]Slagis G.Recommended revisions to nuclear piping thermalexpansion stress limits[J].Journal of Pressure VesselTechnology,Transactions of the ASME,2010,133(2):021102-1-7.
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