• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2003.05a
• /
• pp.82-87
• /
• 2003

Nuclear fuel with a big slenderness ratio is susceptible to flow-induced vibration under very severe conditions of high temperature, high flow and exposure to irradiation in nuclear reactor. The fuel assembly should, therefore, be designed to escape any resonance due to the vibration during the reactor operation, in particular, in case of the design changes. In addition, the amplitudes due to the grid vibration, the fuel rod vibration and the fuel assembly vibration should be minimized to reduce the grid-to-rod fretting wear. Fuel assembly vibration tests in air at room temperature and in water at high temperature have been performed to investigate fuel vibration behaviors. The frequency and damping during the test in air have been compared to those in water. Through the hydraulic test, the advanced assembly has been evaluated not to be susceptible to any resonance. In addition, the test data from the tests can be used to make fuel model and to evaluate grid-to-rod fretting wear.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.35 no.2
• /
• pp.201-206
• /
• 2011

A spacer grid assembly is one of the most important structural components of the nuclear fuel assembly of a pressurized water reactor (PWR), and it affects the performance of the fuel assembly. The primary design requirement is that the mechanical integrity of the fuel rod should be maintained by the spacer grid assembly during the operation of the reactor. It was known that fretting damage to the fuel rod can be reduced by adjusting the relative moving displacement between the fuel rod and its support. In this study, we used the finite element method to evaluate the characteristics of a sliding support designed to reduce fretting damage of fuel rods.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2003.05a
• /
• pp.260-260
• /
• 2003

• Proceedings of the Computational Structural Engineering Institute Conference
• /
• 2006.04a
• /
• pp.828-835
• /
• 2006

Spacer grid springs support the fuel rods in a nuclear fuel system. The spacer grid is a part of a fuel assembly. Since a spring has repeated contacts with the fuel rod, fretting wear occurs on the surface of the spring. Design is usually performed to reduce the wear. The conceptual design process for the spring is defined by using the Independence of axiomatic design and the design is carried out based on the direction that the design matrix indicates. For detailed design an optimization problem is formulated. In optimization, homologous design is employed to reduce fretting wear. The deformation of a structure is called homologous if a given geometrical relationship holds for a given number of structural points before, during, and after the deformation. In this case, the deformed shape of the spring should be the same as that of the fuel rod. 1bis condition is transformed to a function and considered as a constraint in the optimization process. The objective function is minimizing the maximum stress to allow a local plastic deformation. Optimization results show that the contact occurs in a wide range. Also, the results are verified by nonlinear finite element analysis.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2001.05a
• /
• pp.329.1-329
• /
• 2001

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.05a
• /
• pp.274-274
• /
• 1999

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.833-834
• /
• 2004

• Journal of KSNVE
• /
• v.5 no.1
• /
• pp.95-106
• /
• 1995

Flow-induced vibration in heat exchanger (or fuel rod) in nuclar power plant can cause dynamic interactions between tubes and tube supports resulting in fretting-wear. To increase the reliability and design life of heat exchanger components, design criteria that establish acceptable limits of vibration and minimize fretting wear are necessary. The fretting-wear rate is dependent upon material combination, contact configuration, environmental conditions and tube-to tube support dynamic interaction. It is demostrated that the fretting -wear rate correlates well with tube-to-support contact force or work rate. The tube-to-support dynamic interaction, which consists of dynamic contact forces and tube motion, is used to relate single-span wear data to real heat exchanger configurations consisting of multi-span tube bundles. This paper describes the test facility to measure tube-to-support dynamic impact force and reports its dynamic characteristics through the four impact tests - a force transduces independent and external impact tests, central ring inside impact test and additional cylinder impact test. Through the tests the impact parameter change dependent upon the material difference of impacting ball is studied, and the impact parameters of Force Transducer Assembly components are measured. And also the dynamic behavior of Force Transducer Assembly is analyzed. The force measurement technique herein is shown to provide a reasonable measure of dynamic contact forces.

• The Transactions of The Korean Institute of Electrical Engineers
• /
• v.61 no.12
• /
• pp.1948-1952
• /
• 2012

The paper presents modal testing and analysis in order to obtain the dynamic characteristics of a middle spacer grids of a nuclear fuel rod. A spacer grid is one of the important structural elements supporting nuclear fuel rods. Such a fuel rod can be oscillated by its thermal expansion, neutron irradiation and etc. due to cooling water flow under the operation of a nuclear power plant. When the fuel rod vibrates, fretting wear due to repeated friction motion between the fuel rods and spacer grids can be occurred, and so the fuel rod is damaged. In this paper, through modal analysis and testing, natural frequencies and modes of a middle spacer grid were calculated, and the following conclusions were obtained. Firstly the numerical first-seven natural frequencies for spacer grids of a fuel rod having complicated structures have a small difference within 3.8% with experimental natural frequencies, and so the suitability of simulation results was verified. Secondly, experimental mode shapes for a middle spacer grid of a nuclear fuel rod were verified by obtaining lower non-diagonal terms through MAC(Modal Assurance Criteria), and were confirmed by the simulation modes.

• Nuclear Engineering and Technology
• /
• v.33 no.1
• /
• pp.73-82
• /
• 2001

Pressurized water reactor(PWR) fuel rods. which are continuously supported by a spring system called a spacer grid(SG), are exposed to reactor coolant at a flow velocity of up to 6-8 m/s. It is known that the vibration of 3 fuel rod is generated by the coolant flow, a so-called flow-induced-vibration(FIV), and the relative motion induced by the FIV between the fuel rod and the SG can wear away the surface of the fuel rod, which occasionally leads to its fretting failure. It is, therefore, important to understand the vibration characteristics of the fuel rod and reflect that in its design. In this paper, vibration analyses of the fuel rod with two different SGs were performed using both analytical and experimental methods. Updating of the finite element(FE) model using the measured data was performed in order to enhance confidence in the FE model of fuel rods supported by an SG. It was found that the modal parameters are very sensitive to the spring constant of the SG.