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Investigation of the Flow Regime and Flow Induced Vibrations (FIV) of the Control Rod Inside the Guide Channel of PWR

[+] Author Affiliations
Muhammad Ali

Pakistan Atomic Energy Commission, Islamabad, Pakistan

Lu Daogang, Zhang Huimin, Shahroze Ahmed, Muhammad Zeeshan Ali

North China Electric Power University, Beijing, China

Paper No. ICONE25-66855, pp. V006T08A059; 5 pages
doi:10.1115/ICONE25-66855
From:
  • 2017 25th International Conference on Nuclear Engineering
  • Volume 6: Thermal-Hydraulics
  • Shanghai, China, July 2–6, 2017
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-5784-7
  • Copyright © 2017 by ASME

abstract

The phenomenon of Flow induced vibrations (FIV) through the motion induced into the control rod of the nuclear core model apparatus due to high speed internal and external fluid flow. It involves a variety of flow phenomena around the restrictions and over a range of vibration frequencies of the control rod and other related components. The results of a CFD analysis are presented, which are compared with the experimental values obtained from the experimental setup specifically designed for solving and reviewing the fluid induced vibration problems, and thus forming equations of fluid flow regime and vibration modeling. The experimental setup is a model of a nuclear core consists of a fuel assembly without any fuel rod, a control rod held from a rigid screwed support and three fluid mixing plates similar to grid spacers with external and internal fluid flow. There are two inlets, forcing water from below and one of the sides of the apparatus while one outlet located above from both on the opposite side of the second inlet. The results of the experimental setup were detected by laser displacement sensor, to measure the vibration of the control rod, connected with a computer system. This vibration displacement data was measured under different flow conditions rendered due to different pressures through the inlet pumps. The different meshing results give a comparison of 3D modeling under different meshing strategies. Different group of inlet and outlet flow values have been considered according to the designed apparatus including velocities, pressure gradients to form the equations. The procedure followed for numerical method is starting from static fluid flow equations and moving on to the transient equations and thus forming the concluding equation. Similarly, the simulations have been improved from simple steady conditions to the transient solution while for the vibration modeling general modeling rules have been followed and different constants have been taken from the materials libraries available and the values of load from the experimental data and simulation. The results obtained from simulations represent that 3D modeling refinement makes it much complex and as known takes more time and needs heavy computing memories but the results are acceptable to the extent required. So the equations of fluid induced vibrations and fluid flow regime in a vertical nuclear core, by performing simulations of ANSYS CFX and FLUENTS and obtaining the experimental values and calibrating the values with those obtained from simulations as well as numerically solving the problem and finally comparing them all.

Copyright © 2017 by ASME

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