A finite volume approach to the problem of heat transfer in axisymmetric annulus geometry with internal heating element using local analytical solution techniques

被引:1
作者
Salama, A. [1 ]
机构
[1] AEA, Nucl Res Ctr, Cairo 13759, Egypt
关键词
MATERIAL TESTING REACTOR; MTR RESEARCH REACTOR; FLOW BLOCKAGE; COOLANT CHANNEL; CFD ANALYSIS; HOT CHANNEL; SIMULATION; INVERSION;
D O I
10.3139/124.110433
中图分类号
TL [原子能技术]; O571 [原子核物理学];
学科分类号
0827 ; 082701 ;
摘要
In this paper we implement the local analytical solution technique to the problem of heat transfer in axisymmetric annulus geometry with internal heating element. This method has shown to be very accurate in estimating the temperature field for axisymmetric problems even for coarse mesh. It is shown that this method reduces to the analytical solution for unidirectional heat transfer in the radial direction in homogeneous media. The technique is based on finding an analytical expression for the temperature field in the radial direction within each grid cell. This means that the temperature field in each cell is allowed to change in a nonlinear fashion along the radial direction. We compare this technique with the traditional finite volume technique and show that; with only few cells in the radial direction, this technique arrives at the mesh-independent solution quite accurately whereas it required denser mesh to arrive closer to this solution using traditional techniques. This method is proposed to the 1D codes that are currently being used to simulate thermalhydraulic characteristics of reactor systems. Furthermore, we also implement the experimental temperature field algorithm in which the governing equations are approximated for each cell as it would without extra manipulation to the governing equations. This technique is very simple and separates the physics from the solving part.
引用
收藏
页码:436 / 445
页数:10
相关论文
共 20 条
  • [1] Numerical simulation of turbulent flow in a 37-rod bundle
    Chang, D.
    Tavoularis, S.
    [J]. NUCLEAR ENGINEERING AND DESIGN, 2007, 237 (06) : 575 - 590
  • [2] 3D thermal hydraulic simulation of the hot channel of a typical material testing reactor under normal operation conditions
    El-Morshedy, S. El-Din
    Salama, A.
    [J]. KERNTECHNIK, 2010, 75 (05) : 248 - 254
  • [3] A comparative CFD investigation of helical wire-wrapped 7, 19 and 37 fuel pin bundles and its extendibility to 217 pin bundle
    Gajapathy, R.
    Velusamy, K.
    Selvaraj, P.
    Chellapandi, P.
    Chetal, S. C.
    [J]. NUCLEAR ENGINEERING AND DESIGN, 2009, 239 (11) : 2279 - 2292
  • [4] Hamidouche T, 2004, ANN NUCL ENERGY, V31, P1385, DOI 10.1016/j.anucene.2004.03.008
  • [5] Overview of accident analysis in nuclear research reactors
    Hamidouche, Tewfik
    Bousbia-Salah, Anis
    Si-Ahmed, El Khider
    D'Auria, Francesco
    [J]. PROGRESS IN NUCLEAR ENERGY, 2008, 50 (01) : 7 - 14
  • [6] On turbulence models for rod bundle flow computations
    Házi, G
    [J]. ANNALS OF NUCLEAR ENERGY, 2005, 32 (07) : 755 - 761
  • [7] Simulation of loss-of-flow transients in research reactors
    Housiadas, C
    [J]. ANNALS OF NUCLEAR ENERGY, 2000, 27 (18) : 1683 - 1693
  • [8] Thermal-hydraulic modeling of flow inversion in a research reactor
    Kazeminejad, H.
    [J]. ANNALS OF NUCLEAR ENERGY, 2008, 35 (10) : 1813 - 1819
  • [9] Li W., 2011, AS S COMP HEAT TRANS
  • [10] Flow blockage analysis of a channel in a typical material test reactor core
    Lu, Qing
    Qiu, Suizheng
    Su, G. H.
    [J]. NUCLEAR ENGINEERING AND DESIGN, 2009, 239 (01) : 45 - 50