• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.337-338
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• 2005

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3653-3664
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• 2021

In PWR three-dimensional transient coupling calculation software CORCA-K, the nodal Green's function method and diagonal implicit Runge Kutta method are used to solve the spatiotemporal neutron dynamic diffusion equation, and the single-phase closed channel model and one-dimensional cylindrical heat conduction transient model are used to calculate the coolant temperature and fuel temperature. The LMW, NEACRP and PWR MOX/UO₂ benchmarks and FangJiaShan (FJS) nuclear power plant (NPP) transient control rod move cases are used to verify the CORCA-K. The effects of burnup, fuel effective temperature and ejection rate on the control rod ejection process of PWR are analyzed. The conclusions are as follows: (1) core relative power and fuel Doppler temperature are in good agreement with the results of benchmark and ADPRES, and the deviation between with the reference results is within 3.0% in LMW and NEACRP benchmarks; 2) the variation trend of FJS NPP core transient parameters is consistent with the results of SMART and ADPRES. And the core relative power is in better agreement with the SMART when weighting coefficient is 0.7. Compared with SMART, the maximum deviation is -5.08% in the rod ejection condition and while -5.09% in the control rod complex movement condition.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.268-269
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• 2005

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2007.05a
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• pp.197-201
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• 2007

Vibration characteristics of an annular cylindrical fuel rod, which was proposed as a candidate design of fuel's cross section for the ultra-high burnup nuclear fuel, according to the cross-sectional dimensions and the number of supports or the span length were analytically studied. Finite element(FE) modeling for the annular cross sectional fuel was based on the methodology, that have been proven by the test verification, for the conventional PWR nuclear fuel rod. A commercial FEA code, ABAQUS, was used for the FE modeling and analysis. A planar beam element (B21) that uses a linear interpolation was used for the fuel rod and a linear spring element for the spring and dimple of the SG. Natural frequencies and mode shape were calculated according to the preliminary design candidates for the fuel's cross sectional dimension and the number of span. From the analysis results, the design scheme of the annular fuel compatible to the present PWR nuclear reactor core was discussed in terms of the number of supports and fuel's cross section.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.7
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• pp.843-850
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• 2010

A spacer grid assembly is one of the most important structural components of the nuclear fuel assembly of a Pressurized Water Reactor (PWR). A primary design requirement is that the fuel rod integrity be maintained by the spacer grid assembly during the operation of the reactor. In this study, we suggested a new spacer grid assembly having a fuel rod support, which is capable of sliding when the fuel rod vibrates due to flow-induced vibrations in the reactor. By adjusting the relative displacement between the fuel rod and its support, the proposed design will help in reducing fuel rod fretting damage.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 1997.10a
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• pp.177-183
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• 1997

The fuel rod for PWR is supported by the spring of the sapcer grid to maintain its axial location and lateral space between fuel rods to get proper functions during the residence in the reactor. The long exposure duration makes the spring to be relax and loss the spring force that results in a fuel rod rattling which may cause fuel rod failure. Here considering the spring behaviour as a function of burnup the reaction forces of the springs are calculated by the finite element program developed herein to evaluate the integrity of the fuel rod from fretting. The results are compared with previous data and ANSYS for the validation of the program and procedures.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2880-2886
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• 2020

A small leak occurring on the surface of a fuel rod due to damage exposes UO₂ to a steam atmosphere. During this time, fission gas trapped inside the fuel rod leaks out, and the gas leakage can be increased due to UO₂ oxidation. Numerous studies have focused on the steam oxidation and its thermodynamic calculation in UO₂. However, the thermodynamic calculation of the UO₂ oxidation in a pressurized water reactor (PWR) environment has not been studied extensively. Moreover, the kinetics of the oxidation of UO₂ pellet also has not been investigated. Therefore, in this study, the thermodynamics of UO₂ oxidation under steam injection due to a damaged fuel rod in a PWR environment is studied. In addition, the diminishing radius of the UO₂ pellet with time in the PWR environment was calculated through an experiment simulating the initial time of steam injection at the puncture.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2002.05a
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• pp.684-687
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• 2002

The authors have developed two devices which cuts the spend fuel rod to an optimal size and extracts fuel pellet from the pieces of cut fuel rods. These devices are so important to reduce radioactive wastes that some advanced countries developed their own methods and devices. The authors have benchmarked from these methods and devices. For spent fuel rod cutting, the tube cutting method has been chosen. some mechanical properties of the fuel tube and pellet has been carefully considered for an optimal cutting size. For fuel pellet extraction, a mechanically extracting method has been adopted. The existing chemical method have turned out to be inappropriate because it produced large amount of radioactive wastes, in spite of its high fuel recovery characteristics. The developed method has an advantage that it can be applied to other fuel rods that have different shapes and sizes. The two devices are set up and operated in the hot cell where people can not go in, so that the devices have been designed to be controlled remotely and modulated for easy maintenance. And the performance of the devices has been tested by using simulated fuel rod. From the experimental results, the devices are supposed to be useful for reducing radioactive wastes.

• Nuclear Engineering and Technology
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• v.30 no.3
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• pp.273-286
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• 1998

As there has arisen a concern that failure of the high burnup fuel under the reactivity-insertion accident(RIA) may occur at the energy lower than the expected, fuel behavior under the rod ejection accident in a typical Westinghouse-designed 950 MWe PWR was analyzed by using the three dimensional nodal transient neutronics code, PANBOX2 and the transient fuel rod performance analysis code, FRAP-T6. Fuel failure criteria versus the burnup was conservatively derived taking into account available test data and the possible fuel failure mechanisms. The high burnup and longer cycle length fuel loading scheme of a peak rod turnup of 68 MWD/kgU was selected for the analysis. Except three dimensional core neutronics calculation, the analysis used the same core conditions and assumptions as the conventional zero dimensional analysis. Results of three dimensional analysis showed that the peak fuel enthalpy during the rod ejection accident is less than one third of that calculated by the conventional zero dimensional analysis methodology and the fraction of fuel failure in the core is less than 4 %. Therefore, it can be said that the current design limit of less than 10 percent fuel failure and maintaining the core coolable geometry would be adequately satisfied under the rod ejection accident, even though the conservative fuel failure criteria derived from the test data are applied.

• Tribology and Lubricants
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• v.30 no.6
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• pp.343-349
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• 2014

Currently, in order to resist the scratching of the fuel rod surface while fabricating the fuel assembly of the light-water nuclear reactor, we use a solution of nitrocellulose, an explosive material, as a dry lubricant along with its solvent. However, the demand for developing safe and harmless aqueous alternative materials for environment-conservation and field-worker safety has increased. In this study, we demonstrate the preparation of a novel aqueous resin composite using a formulation of aqueous polymeric resin, alcoholic solvent, and water. Subsequently, we characterize this composite on the basis of hardness, adhesive property, and water solubility using plates similar to the fuel rod material. The insertion test of a fuel rod coated with the YS-3 composite shows load values of $18.8-20.5kg/cm^2$, which is comparable with $18.8-20.5kg/cm^2$ of the nitrocellulose coating agent. In addition, the depth and width of longitudinal scratches caused by the YS-3 composite test are 50% higher than those of the standard. We can develop a harmless and safe aqueous dry lubricant to replace the existing NC products through field testing of 264 pieces of fuel rods, after producing 350 kg of the YS-3 prototype. The scratch test for the rod surface showed that weight of chip of YS-3 prototype was smaller than that of NC before and after solvent treatment, indicating the properties of YS-3 prototype was comparable to the counterpart.