• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.424-429
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• 1993

Numeric characters are printed at the end part of nuclear fuel rod containing nuclear pellets. Fuel rods are discriminated and managed systematically by these characters in the process of producing fuel assembly. The characters are also used to examine manufacturing process of fuel rods in the survey of burnup efficiency as well as in inspection of irradiated fuel rod. Therefore automatic character recognition is one of the most important technologies in automatic manufacture of fuel assembly. In this study, character recognition system is developed. In the developed system, mesh feature extracted from each character written in the fuel rod has been compared with reference feature value stored in database, and the character is thus identified. In the result of experiment, 95.83 percent recognition rate is achievable.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.04a
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• pp.781-782
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• 2012

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2006.05a
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• pp.246-249
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• 2006

Finite element vibration analysis of the trial 5x5 partial fuel assembly in the still water was performed using the substructure method. ANSYS software was used as a finite element modeling and modal analysis tool. The calculated natural frequencies of the partial fuel assembly were more consistent with the experimental results for the identical test model compared to the much larger solid model. This modeling technique can be utilized for the fuel assembly dynamic behavior analysis under normal operation, seismic and loss-of-coolant-accident analysis.

• Nuclear Engineering and Technology
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• v.21 no.3
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• pp.193-204
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• 1989

This paper develops a new method for reconstructing neutron flux distribution, that is based on the maximum entropy Principle in information theory. The Probability distribution that maximizes the entropy Provides the most unbiased objective Probability distribution within the known partial information. The partial information are the assembly volume-averaged neutron flux, the surface-averaged neutron fluxes and the surface-averaged neutron currents, that are the results of the nodal calculation. The flux distribution on the boundary of a fuel assembly, which is the boundary condition for the neutron diffusion equation, is transformed into the probability distribution in the entropy expression. The most objective boundary flux distribution is deduced using the results of the nodal calculation by the maximum entropy method. This boundary flux distribution is then used as the boundary condition in a procedure of the imbedded heterogeneous assembly calculation to provide detailed flux distribution. The results of the new method applied to several PWR benchmark problem assemblies show that the reconstruction errors are comparable with those of the form function methods in inner region of the assembly while they are relatively large near the boundary of the assembly. The incorporation of the surface-averaged neutron currents in the constraint information (that is not done in the present study) should provide better results.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.16 no.2 s.107
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• pp.163-168
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• 2006

The PWR Nuclear Fuel assembly consists of more than 250 fuel rods that are supported by leaf springs in the cells of more than 10 Spacer Grids (SG) along the rod length. Since it is not easy to conduct mechanical tests on a full-scale model basis, the small-scaled rod bundle $(5\times5)$ which is called partial fuel assembly is generally used for various performance tests during the development stage. As one of the small-scaled tests, a flow test should be carried out in order to verify the performance of the spacer grid to obtain the Flow-Induced Vibration (FIV) characteristics of the scaled fuel assembly over the specified flow range. A vibration test should be also performed to obtain the modal parameters of the assembly prior to the flow test. In this study, we want to develop the estimation procedure of the damping ratio for the scaled test assembly. For the damping factor of the partial fuel assembly and the grid cage at the first vibration mode, as one of the vibration tests, a so-called pluck testing has been performed in air as a preliminary test prior to in-flow damping measurement test. Logarithmic decrement method is used for calculation of the damping ratio. Estimated damping ratio of the partial fuel assembly is about $0.7\%$ with reasonable error of $2\%$ for the previous results. Nonlinear behavior of the partial fuel assembly might be stem mainly from the rod-grid support configuration.

• Journal of Radiation Protection and Research
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• v.31 no.1
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• pp.37-46
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• 2006

For the first step to develop a reference disposal container of spent fuel to be used in a deep geological repository, this paper examined safe dimensions of the disposal container on the points of nuclear criticality and radiation safety and mechanical structural safety and provided basic information for dimensioning the container and configuration of the container components, and establishing the favorable and safe disposal conditions. When the safety factor for stress due to the external loads (hydrostatic and swelling pressure) is taken as 2.0, the safe diameter of the filler material to provide enough container strength under the assumed external loads is found to be 112cm with 13cm spacing between inner baskets in PWR container. Also the thickness of the thinner section between the fuel basket and the surface of the cast insert is determined to be 150 mm. Regarding these dimensions of the container, the PWR fuel container is sketched to accommodate 4 square assemblies or 297 CANDU fuel 297 bundles (33 circle tubes x 9 stacks). However the top and bottom parts need to be checked again through the detail radiation shielding analysis with respects to the emplacement position and handling processes of the disposal container.

• Journal of Radiation Protection and Research
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• v.35 no.1
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• pp.21-25
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• 2010

Burnup of spent fuel should be determined accurately for the safety control of spent fuel. Especially, it is necessary to measure the burnup profile along the nuclear fuel axis. In the present work, an ionization chamber was designed and fabricated to measure the gamma ray profile inside the guide tube of spent fuel. The ionization chamber was composed of three parts; induction part, gas-inlet part, and sensor part. The sensor part had two electrodes; cathode and anode. A guide electrode was considered in the ionization chamber design to make the ionization chamber to be inserted easily into the guide tube. Pure gas (argon and xenon) was inserted into the ionization chamber, and the leakage current and saturation curve were measured to determine the operation characteristics of the ionization chamber. The gamma ray radiation was also measured in relatively high dose environment. The gamma ray profile of the spent fuel will be measured with the ionization chamber.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.04a
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• pp.911-916
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• 2012

It is important to identify dynamic characteristics of nuclear fuel components. Since the fuel always exposed to turbulent flow, the dynamic contact between grids and rods is one of the fuel failure modes. The dynamic behavior of grids in nuclear fuels is quite complex, since two pairs of spring support are placed in the limited space. The strap in a cell has single spring and double dimples and this paper focuses on investigation of the grid strap(Test Fuel Strap, TFS) vibration in one cell. To identify the grid strap vibration, modal analysis of the strap is performed using Finite Element Method (FEM). Modal testing on a $5{\times}5$ grid structure without rods is performed. The modal testing results are compared to analytic results. In addition, random test considering rod effect is performed about a $5{\times}5$ grid with rods under real contact condition in the air. Finally, the strap vibration of a $5{\times}5$ fuel bundle in INvestigation of Flow INduced vIbraTion(INFINIT) facility is measured in real fluid velocity condition without heating. It is shown that modal frequencies from the test are almost equal to those peak frequencies in the INFINIT test.