l Yield behavior of porous nuclear fuel (UO2)

被引:2
|
作者
Aravindan, S. [1 ,2 ]
Jalaldeen, S. [2 ]
Chellapandi, P. [2 ]
Swaminathan, N. [1 ]
机构
[1] Indian Inst Technol, Dept Mech Engn, Room 206, Madras 600036, Tamil Nadu, India
[2] Indira Gandhi Ctr Atom Res, Reactor Design Grp, Kalpakkam 603102, Tamil Nadu, India
关键词
Uranium dioxide; porous material; finite element method; yield surface; concrete damaged plasticity; FISSION-GAS RELEASE; PELLET-CLADDING INTERACTION; SINTERED URANIUM-DIOXIDE; PLASTIC-DAMAGE MODEL; ELASTIC PROPERTIES; ULTRASONIC VELOCITY; CONCRETE STRUCTURES; INTERNAL-PRESSURE; POPULATIONS; SIMULATION;
D O I
10.1080/15376494.2015.1059529
中图分类号
T [工业技术];
学科分类号
08 ;
摘要
Uranium dioxide (UO2) is one of the most common nuclear fuels. During burn-up, the fuel undergoes substantial microstructural changes including the formation of pressurized pores, thus becoming a porous material. These pores reduce the elastic modulus and alter the yield behavior of the material. In this work, a finite-element-based homogenization technique has been used to map the yield surface of UO2 with pressurized pores. Two scenarios are considered; in the first, the fuel matrix is a ductile material with a Von-mises type behavior, while in the second, the matrix is quasi brittle, which is simulated using the concrete damaged plasticity (CDP) model available in ABAQUS. For both of the scenarios, it is found that the yield strength decreases with an increase in porosity for a given internal pore pressure. For a given porosity, the yield surface shifts towards the negative hydrostatic axis in the Haigh-Westergard stress space with an increase in pore pressure. When the matrix is quasi brittle, the decrease in tensile hydrostatic strength is less than the increase in compressive hydrostatic strength, whereas in the case of a ductile matrix, the changes in the hydrostatic strengths are same. Furthermore, the shape of the yield surface changes from one deviatoric plane to another in both scenarios. Analytical equations, which are functions of pore pressure and porosity, are developed to describe the yield surface of porous UO2 while accounting for the changes in shape of the yield surface from one deviatoric plane to another. These yield functions can be used to predict the failure of porous UO2 fuel.
引用
收藏
页码:1149 / 1162
页数:14
相关论文
共 50 条
  • [31] Peculiar Thermal Behavior of UO2 Local Stucture
    Prieur, Damien
    Epifano, Enrica
    Dardenne, Kathy
    Rothe, Joerg
    Hennig, Christoph
    Scheinost, Andreas C.
    Neuville, Daniel R.
    Martin, Philippe M.
    INORGANIC CHEMISTRY, 2018, 57 (23) : 14890 - 14894
  • [32] BEHAVIOR OF NB2O5 DOPED UO2 FUEL IN REACTIVITY INITIATED ACCIDENT CONDITIONS
    YANAGISAWA, K
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 1991, 28 (05) : 459 - 471
  • [33] Models for fuel porosity evolution in UO2 under various regimes of reactor operation
    Tarasov, V. I.
    Veshchunov, M. S.
    NUCLEAR ENGINEERING AND DESIGN, 2014, 272 : 65 - 83
  • [34] Investigation of burnup and temperature effects on the grain size of UO2 fuel along fission gas behavior
    Sadeghnoedoost, A.
    Zolfaghari, A.
    Shirani, A. S.
    NUCLEAR ENGINEERING AND DESIGN, 2025, 435
  • [35] Athermal dislocation strengthening in UO2
    Portelette, Luc
    Amodeo, Jonathan
    Michel, Bruno
    Madec, Ronan
    JOURNAL OF NUCLEAR MATERIALS, 2020, 538
  • [36] On the condition of UO2 nuclear fuel irradiated in a PWR to a burn-up in excess of 110 MWd/kgHM
    Restani, R.
    Horvath, M.
    Goll, W.
    Bertsch, J.
    Gavillet, D.
    Hermann, A.
    Martin, M.
    Walker, C. T.
    JOURNAL OF NUCLEAR MATERIALS, 2016, 481 : 88 - 100
  • [37] A phase field study of the thermal migration of gas bubbles in UO2 nuclear fuel under temperature gradient
    Wang, Yafeng
    Xiao, Zhihua
    Hu, Shenyang
    Li, Yulan
    Shi, San-Qiang
    COMPUTATIONAL MATERIALS SCIENCE, 2020, 183
  • [38] Porosity evolution study in irradiated UO2 fuel based on fuel matrix swelling
    Roostaii, B.
    Kazeminejad, H.
    Khakshournia, S.
    KERNTECHNIK, 2018, 83 (02) : 86 - 90
  • [39] Impact of thermal conductivity models on the coupling of heat transport and oxygen diffusion in UO2 nuclear fuel elements
    Mihaila, Bogdan
    Stan, Marius
    Crapps, Justin
    JOURNAL OF NUCLEAR MATERIALS, 2012, 430 (1-3) : 221 - 228
  • [40] Modelling of irradiated UO2 fuel behaviour under transient conditions
    Veshchunov, M. S.
    Tarasov, V. I.
    JOURNAL OF NUCLEAR MATERIALS, 2013, 437 (1-3) : 250 - 260