• Proceedings of the KSME Conference
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• 2003.11a
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• pp.1297-1302
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• 2003

KAERI has contrived 15 kinds of spacer grid shapes of its own since 1997 and applied for domestic and foreign patents. To date, KAERI has obtained US and ROK patents for 6 kinds of spacer grid shapes among them and the others are under review in USA, EC, China, and ROK. In this study, mechanical/structural performance analysis and test on two spacer grid shapes that are assumed to be the most effective candidates for the spacer grid of the next generation nuclear fuel in Korea was carried out. The result has shown that the performances of the candidates are better or not worse than those of the current spacer grid.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.1819-1824
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• 2004

The common method to improve heat transfer in Nuclear fuel rod bundle is install a mixing vane in space grid. The previous split mixing vane is guides cooling water to swirl flow in sub-channel of fuel assembly. But, this swirl flow decade rapidly after mixing vane and the effect of enhancing the heat transfer vanish behind this short region. The large scale secondary vortex flow was generated by rearranging the inclined angle direction of mixing vanes to the coordinated directions. This LSVF mixing vanes generate the most strong secondary flow vortices which maintain about 35 $D_H$ after the spacer grid and the streamwise vorticity in subchannel with LSVF mixing vane sustain two times more than that in subchannel with split mixing vane. The turbulent kinetic energy and the Reynolds stresses generated by the mixing vanes have nearly same scales but maintain twice more than previous type.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.432-437
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• 2004

KAERI has contrived 16 kinds of spacer grid shapes of its own since 1997 and applied for domestic and foreign patents. To date, KAERI has obtained US and ROK patents for 11 kinds of spacer grid shapes among them and the others are under review in USA, EC, China, and ROK. In this study, detailed performance analysis and test on two spacer grid shapes that are assumed to be the most effective candidates for the spacer grid of the next generation nuclear fuel in Korea was carried out. The result has shown that the performances of the candidates are better or not worse than those of the current spacer grid.

• Nuclear Engineering and Technology
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• v.13 no.2
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• pp.63-72
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• 1981

The change in the design-basis refueling strategy caused by the unexpected nuclear fuel failures may result in discharging intact fuel assemblies which were irradiated in the positions symmetric to the failed ones in addition to the failed ones in order to maintain the symmetric power shape in the reactor core. In this work an attempt is made to reuse the intact fuel assemblies which were discharged before reaching the design turnup in the above-described situation so as to improve the fuel utilization. The TDCORE code is used to estimate the flux and power distribution, and the RELOAD-II code for searching the optimal loading pattern with the minimum assembly radial power peaking factor. For the case of the Ko-ri unit 1, its third cycle turnup could be extended to 11,648 MWD/MTU by reusing the four low-burned fuel assemblies removed at the end of the first cycle, and then the loading pattern is searched to the equilibrium cycle.

• Nuclear Engineering and Technology
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• v.22 no.2
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• pp.95-100
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• 1990

The exact expression for the 1151 force on a fuel assembly in a reactor core is derived in terms of calculable hydraulic parameters. The relation for the lift force. pressure drop, buoyancy force, viscous force. and fuel assembly weight is discussed. Based on the derived exact expression. error analysis is made for a simple expression applying COBRA IV-i to a typical PWR fuel assembly. The error analysis revealed that the error of the simple expression consists of four terms and the overall error depends on the flow rate change direction, and its magnitude is about 1%.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.20-27
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• 1993

This paper concerns with developing a simplified in-core fuel management scoping tool for PWR. For this purpose the point reactivity model is put into a fuel cycling decision code, FCYPRM. Modified Borresen's coarse-mesh diffusion theory and nodal expansion method are utilized to form a spatial neutron analysis code, CMSNAP. Numerical experiments are per- formed to determine a set of empirical shuffling rules for working out an automated fuel loading pattern search code, ALPS. The numerical examples are presented for verifying effectiveness and applicability of individual codes. By structuring and applying three codes for reload core design problem of a PWR, it is demonstrated that these codes provide an effective in-core fuel management scoping tool for PWR.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.43-49
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• 2022

The purpose of this study is to evaluate the radiation effect on damage when the external shield of the spent nuclear fuel transport cask is damaged due to impact as the cause of an unexpected accident. The neutron and gamma-ray intensities and spectra are calculated using the ORIGEN-Arp module in the SCALE 6.2.4 code package⁽¹⁾ and then using MCNP6.2⁽²⁾ code calculate the dose rate. In order to evaluate the radiation dose according to the size of damage caused by external impact, various sized holes of 0.3~13.7% are assumed in the outer shield of the cask to evaluate the sensitivity to the dose. In the case of radiation source leakage, damage to the nuclear fuel assembly is assumed to be up to 6% based on overseas test cases. When only the outer shield is damaged, the maximum surface dose is calculated as 3.12E+03 mSv/hr. However, if the radiation source is leaked due to damage to the nuclear fuel assembly, it becomes 7.00E+05 mSv/hr which is about 200 times greater than the former case.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2008.11a
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• pp.346-347
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• 2008

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2006.11a
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• pp.286-286
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• 2006

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2007.11a
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• pp.273-274
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• 2007