• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.66-74
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• 1990

The thermal-hydraulic characteristics of the 17$\times$17 OFA (Optimized Fuel Assembly) used in the KNU 7&8 are analyzed and compared with that of the 17$\times$17 SFA (Standard Fuel Assembly) loaded in the KNU 5&6. The thermal-hydraulic characteristics analyzed are minimum DNBR, fuel centerline temperature and exit void fraction at normal operation and design over power transient. Additionally, local linear rod power, which will cause fuel centerline melting, is calculated. The DNBR sensitivity calculations are performed with respect to the reactor operating parameters. COBRA-IV-I code is used for these calculations. The modified W-3 correltion and the drift-flux model are applied for the critical heat flux calculation and the void fraction calculation, respectively. From the calculated results, it has been found that the possibility of DNB occurrence is higher in the OFA than in the SFA. The other hand, the local linear power resulting in fuel centerline moiling of the OFA is nearly equal to that of the SFA.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2007.11a
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• pp.439-440
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• 2007

• Transactions of the Korean Society of Mechanical Engineers
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• v.8 no.1
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• pp.48-56
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• 1984

In order to estimate the fuel rod temperature during the reflooding phase of the PWR LOCA, it is essential to obtain a better understanding of the rewetting mechanism. This paper presents the results of analytical and experimental investigations aimed at analyzing the rewetting phenomena in a heated tube. A two-dimensional solution of the rewetting for a tube is described and used to yield the correlation of the rewetting heat transfer coefficient as the function of flooding rate and inlet subcooling. This correlation prediction is in good agreement with the experimental data.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.879-880
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.945-946
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• 2004

• Journal of the Korean Society for Nondestructive Testing
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• v.9 no.2
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• pp.8-17
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• 1989

Reconstruction of the radial two-dimensional of fission products using projections obtained by nondestructive transversal direction gamma scanning was evaluated, and then compared with the destructive sectional gamma scanning results. Although the nondestructive method has a little drawback compared to the destructive method in the accuracy, the former seemed to be sufficiently applicable.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.234-234
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• 2002

• Journal of Welding and Joining
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• v.10 no.4
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• pp.240-249
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• 1992

The objective of this study is to investigate the characteristics of welded joints on the Zircaloy-4 resistance upset welding for HWR(Heavy Water reactor)fuel rods. To estimate the characteristics of welded joints, the various tests were performed on the test coupons systematically with a wide range of each welding parameters in terms of a tensile test, burst test, knoop hardness test and metallography. Major results obtained in this study are as follows: 1. The tube and machined with 120.deg. projection was the reliable weld joint design for the nuclear fuel rod end cap welding. 2. As the weld current and the amount of upset increased linearly with increasing welding main heat input, it could make an estimate of their variation in accordance with the phase shift control. 3. It was found that an increase in squeeze force has an effect on the upset contour of welded joint because the amount of upset were increased by the change of squeeze force.

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.47-50
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• 2002

The numerical simulation of a 271-rod fuel assembly of nuclear Liquid-Metal Fast Reactor (LMFR) with an infernal blockage has been carried out. Internal blockage within a subassembly is addressed in the safety assessment because it potentially has very serious consequences for the reactor as a whole. Three dimensional calculations were performed using the SABRE4 computer code for the range of blockage positions and sizes to investigate the seriousness and detectability of the internal blockage. The magnitude and location of the peak temperatures together with the temperature distribution at the subassembly exit were calculated in order to look at the potential for damage within the subassembly, and the possibility of blockage detection. The analysis result shows that the 6-subchannel blockage causes large temperature rise within a assembly with practically no change in mixed mean temperature at the assembly exit.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.57-61
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• 2008

During the period of retubing work for the licensing renewal, the fuel channels, calandria tubes and feeders of CANDU Reactors will be replaced, and calandria visual examination will be performed. This period is a unique opportunity to inspect the inside of the calandria. The visual inspection for the calandria vessel and its internals of Wolsong NPP is scheduled for confirming the calandria integrity. The first visual inspection for the calandria is planned in Pt. Lepreau led by AECL. The visual inspection for Wolsong NPP, led by NETEC(Nuclear Engineering & Technology Institute) of KHNP, will employ 3D laser scanner and 3D CAD Mock-up for the first time in the world, in addition to a conventional video camera. The inspection system is composed of a robot with the 3D laser scanner, a video camera and a hardness meter.