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预应力安全壳在热-压耦合作用下的易损性分析

杨青屿 马燕 严佳川 高戈 刘蒙莎

杨青屿, 马燕, 严佳川, 高戈, 刘蒙莎. 预应力安全壳在热-压耦合作用下的易损性分析[J]. 核动力工程, 2025, 46(4): 181-191. doi: 10.13832/j.jnpe.2024.070048
引用本文: 杨青屿, 马燕, 严佳川, 高戈, 刘蒙莎. 预应力安全壳在热-压耦合作用下的易损性分析[J]. 核动力工程, 2025, 46(4): 181-191. doi: 10.13832/j.jnpe.2024.070048
Yang Qingyu, Ma Yan, Yan Jiachuan, Gao Ge, Liu Mengsha. Fragility Analysis of Prestressed Containment under Thermal-Compressive Coupling Condition[J]. Nuclear Power Engineering, 2025, 46(4): 181-191. doi: 10.13832/j.jnpe.2024.070048
Citation: Yang Qingyu, Ma Yan, Yan Jiachuan, Gao Ge, Liu Mengsha. Fragility Analysis of Prestressed Containment under Thermal-Compressive Coupling Condition[J]. Nuclear Power Engineering, 2025, 46(4): 181-191. doi: 10.13832/j.jnpe.2024.070048

预应力安全壳在热-压耦合作用下的易损性分析

doi: 10.13832/j.jnpe.2024.070048
基金项目: 国家重点研发计划资助项目(2020YFB1901403)
详细信息
    作者简介:

    杨青屿(1997—),男,博士生,现主要从事安全壳结构性能研究工作,E-mail: 18821787236@163.com

    通讯作者:

    严佳川,E-mail: yanjiachuan@hit.edu.cn

  • 中图分类号: TL353

Fragility Analysis of Prestressed Containment under Thermal-Compressive Coupling Condition

  • 摘要: 本文对预应力安全壳结构进行了建模,利用有限元分析软件ABAQUS模拟热-压耦合试验,利用拉丁超立方抽样法得到的安全壳随机样本进行计算,得到两条安全壳易损性曲线,分析安全壳整体功能性失效、结构性失效对应的易损性。计算结果表明,安全壳的下限和上限内压承载力分别是0.9666 MPa和1.0352 MPa。钢衬里功能性失效准则下,HRB400钢筋弹性模量对安全壳的内压承载力影响最大;钢衬里最大拉应变集中分布在设备闸门洞口附近。预应力筋结构性失效准则下,HRB500钢筋弹性模量对安全壳的内压承载力影响最大;预应力筋最大拉应变的分布没有明显的规律。

     

  • 图  1  安全壳有限元模型

    Figure  1.  Finite Element Model of Containment

    图  2  有限元模拟方法验证[23]

    Figure  2.  Validation of the Finite Element Simulation Method[23]

    图  3  温度和压力加载曲线

    Figure  3.  Temperature and Pressure Loading Curves

    图  4  沿筒壁厚度方向节点温度

    Figure  4.  Nodal Temperature along the Cylinder Wall Thickness Direction

    图  5  混凝土受拉损伤

    Figure  5.  Concrete Tensile Damage

    图  6  钢衬里拉应变

    Figure  6.  Steel Lining Tensile Strain

    图  7  混凝土开裂时安全壳结构变形图

    Figure  7.  Deformation Diagram of Containment Structure during Concrete Cracking

    图  8  不同失效准则对应的安全壳结构变形图

    Figure  8.  Deformation Diagram of Containment Structure under Different Failure Criteria

    图  9  热-压耦合作用下安全壳内压承载力分布

    Figure  9.  Distribution of Pressure Bearing Capacity in Containment under Thermal-compressive Coupling Condition

    图  10  安全壳在热-压耦合作用下的易损性曲线

    Figure  10.  Fragility Curve of Containment under Thermal-compressive coupling Condition

    图  11  部分钢衬里在达到功能性失效应变阈值时的应变图

    Figure  11.  Strain Diagrams of Some Steel Liner at the Time of Functional Failure

    图  12  部分预应力筋在达到结构性失效应变阈值时的应变图

    Figure  12.  Strain Diagrams of Partially Prestressed Tendons at the Time of Functional Failure

    表  1  随机变量概率统计特性

    Table  1.   Probabilistic and Statistical Properties of Random Variables

    随机变量分布类型均值/MPa变异系数
    混凝土抗压强度对数正态460.10
    预应力筋弹性模量正态2000000.03
    预应力筋抗拉强度对数正态18600.07
    HRB400钢筋弹性模量正态2000000.03
    RB400钢筋抗拉强度对数正态4350.07
    HRB500钢筋弹性模量正态2000000.03
    HRB500钢筋抗拉强度对数正态5400.07
    钢衬里弹性模量正态2000000.03
    钢衬里屈服强度对数正态3200.07
    下载: 导出CSV

    表  2  荷载施加步骤

    Table  2.   Load Application Procedure

    步骤 内压/MPa 壳内温度/℃
    1(施加重力荷载) 0 0
    2(进行预应力张拉) 0 0
    3 0.10 40
    4 0.30 56.7
    5 0.45 90
    6 0.45 140
    7 0.52 145
    8 0.60 145
    9  每个荷载步增加0.1 MPa。内压达到1.3 MPa前温度维持在145℃,在1.3 MPa后温度维持在160℃
    直到破坏
    下载: 导出CSV

    表  3  热传导模型材料参数

    Table  3.   Material Parameters of Heat Transfer Model

    材料 线膨胀
    系数/℃−1
    比热容
    /[J ·(kg·℃)−1]
    热传导系数
    / [W·(m·℃)−1]
    热对流系数
    / [W·(m2·K)−1]
    混凝土 1.0×10−5 960 1.355 16
    钢衬里 1.2×10−5 470 45 8
    下载: 导出CSV

    表  4  热传导分析模型单元类型

    Table  4.   Types of Units in Heat Transfer Analysis Model

    单元编号单元类型模拟部分单元数量/个结点数量/个
    DC3D8热传导单元混凝土18984302176198
    DS4R热传导单元钢衬里21307共84460
    DS3热传导单元钢衬里1264
    下载: 导出CSV

    表  5  安全壳易损性曲线参数

    Table  5.   Containment Fragility Curve Parameters

    失效准则 pm/MPa β
    钢衬里功能性失效准则 1.0005 0.0208
    预应力筋结构性失效准则 1.4099 0.0663
    下载: 导出CSV

    表  6  安全壳不同失效准则对应5%、95%分位内压承载力

    Table  6.   Different Failure Criteria of Containment Corresponding to 5% and 95% Internal Pressure Bearing Capacity

    失效准则5%分位内压
    承载力/MPa
    95%分位内压
    承载力/MPa
    钢衬里功能性失效准则0.96661.0352
    预应力筋结构性失效准则1.30351.5215
    下载: 导出CSV

    表  7  两种失效准则随机变量的归一化敏感度指数

    Table  7.   Normalized Sensitivity Index of Random Variables under Two Failure Criteria

    随机变量 归一化敏感度指数
    钢衬里功能性
    失效准则
    预应力筋结构性
    失效准则
    混凝土抗压强度 −0.00464 0.01396
    预应力筋弹性模量 −0.30079 −0.03254
    预应力筋抗拉强度 −0.06129 0.08001
    HRB400钢筋弹性模量 −0.33510 0.48352
    HRB400钢筋抗拉强度 0.08307 −0.10999
    HRB500钢筋弹性模量 −0.05173 0.73778
    HRB500钢筋抗拉强度 −0.01596 −0.26238
    钢衬里弹性模量 0.15805 0.08134
    钢衬里屈服强度 0.01535 −0.05939
    下载: 导出CSV
  • [1] 鲁刚,郑宽. 能源高质量发展要求下核电发展前景研究[J]. 中国核电,2019,12(5):498-502.
    [2] SKILLMAN G R, REMPE J L. The Three Mile Island Unit 2 Accident [M]//GREENSPAN E. Encyclopedia of Nuclear Energy. Oxford: Elsevier. 2021: 17-29.
    [3] SICH A R. The chernobyl nuclear power plant Unit-4 accident[M]//GREENSPAN E. Earth Systems and Environmental Sciences. Amsterdam: Elsevier, 2021: 30-52.
    [4] OHBA T, TANIGAWA K, LIUTSKO L. Evacuation after a nuclear accident: Critical reviews of past nuclear accidents and proposal for future planning[J]. Environment International, 2021, 148: 106379. doi: 10.1016/j.envint.2021.106379
    [5] HORSCHEL D S, CLAUSS D B. The response of steel containment models to internal pressurization[C]//Structural Engineering in Nuclear Facilities. New York, NY: ASCE, 1984: 534-553.
    [6] HESSHEIMER M F, KLAMERUS E W, LAMBERT L D, et al. Overpressurization test of a 1: 4-scale prestressed concrete containment vessel model: NUREG/CR-6810[R]. Albuquerque: Sandia National Laboratories, 2003.
    [7] HESSHEIMER M F, MATHET E. CSNI international standard problem (ISP): NEA/CSNI/R(2005)5[R]. Paris: OECD Nuclear Energy Agency, 2005.
    [8] GUPTA R, ROSSAT D, DÉROBERT X, et al. Blind comparison of saturation ratio profiles on large RC structures by means of NDT and SFE—Application to the VeRCoRs mock-up[J]. Engineering Structures, 2022, 258: 114057. doi: 10.1016/j.engstruct.2022.114057
    [9] CHARPIN L, NIEPCERON J, CORBIN M, et al. Ageing and air leakage assessment of a nuclear reactor containment mock-up: VERCORS 2nd benchmark[J]. Nuclear Engineering and Design, 2021, 377: 111136. doi: 10.1016/j.nucengdes.2021.111136
    [10] 薛荣军,王洪良,褚濛,等. 预应力安全壳内压作用下的有限元研究[J]. 建筑结构,2018, 48(8): 77-82.
    [11] WU J Y, HAO D D, LI W S, et al. Numerical modeling and simulation of a prestressed concrete containment vessel[J]. Annals of Nuclear Energy, 2018, 121: 269-283. doi: 10.1016/j.anucene.2018.06.039
    [12] YAN J H, LIN Y Z, WANG Z F, et al. Failure mechanism of a prestressed concrete containment vessel in nuclear power plant subjected to accident internal pressure[J]. Annals of Nuclear Energy, 2019, 133: 610-622. doi: 10.1016/j.anucene.2019.07.013
    [13] WANG Z F, YAN J C, LIN Y Z, et al. Study on failure mechanism of prestressed concrete containments following a loss of coolant accident[J]. Engineering Structures, 2020, 202: 109860. doi: 10.1016/j.engstruct.2019.109860
    [14] 周磊,钟红,李建波,等. 按不同标准设计的CPR1000安全壳内压易损性分析[J]. 工业建筑,2017, 47(1): 16-20.
    [15] ZHOU L, LI J B, ZHONG H, et al. Fragility comparison analysis of CPR1000 PWR containment subjected to internal pressure[J]. Nuclear Engineering and Design, 2018, 330: 250-264.
    [16] 金松,李忠诚,蓝天云,等. 严重事故下预应力混凝土安全壳非线性分析及性能评估[J]. 核动力工程,2020, 41(4): 96-100.
    [17] 金松,李鑫波,贡金鑫.严重事故下核电厂安全壳结构概率性能评价[J].工程力学, 2021, 38(06): 103-112.
    [18] JIN S, LI Z C, DONG Z F, et al. A simplified fragility analysis methodology for containment structure subjected to overpressure condition[J]. International Journal of Pressure Vessels and Piping, 2020, 184: 104104. doi: 10.1016/j.ijpvp.2020.104104
    [19] JIN S, GONG J X. Fragility analysis and probabilistic performance evaluation of nuclear containment structure subjected to internal pressure[J]. Reliability Engineering & System Safety, 2021, 208: 107400.
    [20] MANDAL T K, GHOSH S, PUJARI N N. Seismic fragility analysis of a typical Indian PHWR containment: comparison of fragility models[J]. Structural Safety, 2016, 58: 11-19. doi: 10.1016/j.strusafe.2015.08.003
    [21] Joint Committee on Structural Safety (JCSS). Probabilistic Model Code: Part 3 – Material Properties:No. JCSS-2001 [R]. Lyngby, Denmark: JCSS, 2001.
    [22] YAO D, GAO G, YANG Q, et al. Mechanical behaviors and internal pressure bearing capacity of nuclear containment using UHPC and ECC: From numerical simulation, machine learning prediction to fragility analysis [J]. Nuclear Engineering and Design, 2024, 429: 113617.
    [23] YANG Q Y, YAN J C, FAN F. Pretest analysis of a prestressed concrete containment 1: 3.2 scale model under thermal-pressure coupling conditions[J]. Nuclear Engineering and Technology, 2023, 55(6): 2069-2087. doi: 10.1016/j.net.2023.03.001
    [24] U. S. Nuclear Regulatory Commission. Containment structural integrity evaluation for internal pressure loadings above design-basis pressure Regulatory guide: Regulatory Guide 1.216[R]. US: U. S. Nuclear Regulatory Commission, 2010.
    [25] HESSHEIMER M F, DAMERON R A. Containment integrity research at Sandia national laboratories-an overview: NUREG/CR-6906[R]. Washington: Division of Fuel, Engineering & Radiological Research, Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, 2006.
    [26] 杨青屿. 某预应力安全壳热-压作用下破坏机理分析及试验设计[D]. 哈尔滨: 哈尔滨工业大学,2021. doi: 10.27061/d.cnki.ghgdu.2021.003633.
    [27] JIN S, LI Z C, LAN T Y, et al. Fragility analysis of prestressed concrete containment under severe accident condition[J]. Annals of Nuclear Energy, 2019, 131: 242-256.
    [28] 马燕. 某预应力安全壳热-压耦合作用下易损性分析[D]. 哈尔滨: 哈尔滨工业大学,2022.
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出版历程
  • 收稿日期:  2024-07-24
  • 修回日期:  2024-10-06
  • 刊出日期:  2025-08-15

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