• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.463-469
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• 2003

PWR-GALE Code is a computerized mathematical model for calculating the releases of radioactive material in gaseous and liquid effluents from PWRs. In PWR-GALE Code, Auxiliary building iodine removal efficiency, one of the code input data, did not reflect adequately the new design of KSNP which has two auxiliary buildings(PAB and SAB). In this study, we developed a revised method how to correct iodine removal efficiency in KSNP. And newly proposed methodology through case study using Ul-Jin 5, 6 design data was verified.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.241.2-241.2
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• 2003

• Proceedings of the KSME Conference
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• 2003.11a
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• pp.1358-1363
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• 2003

Due to the application of Integrated Head Assembly (IHA) in KSNP+ reactor design, an investigation of reactor internals integrity is carried out to assure that the adoption of IHA does not affect the safety of reactor operation. One of the postulated accident events is the R.V. closure head fall from 5.5m high directly above the reactor vessel that may occur during the refueling operation. The analysis model consists of lumped mass elements of the entire reactor vessel and internals. Because of extreme load, separate elastic-plastic analyses are done for the members that undergo plastic deformation. The analysis verified that the stresses of the reactor internals and the fuel assemblies are within the bound of allowable stress limits and the integrity of the fuel assemblies is maintained.

• Journal of Mechanical Science and Technology
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• v.18 no.1
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• pp.74-81
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• 2004

The KSNP Steam Generators (Youngkwang Unit 3 and 4, Ulchin Unit 3 and 4) have a problem of U-tube fretting wear due to Flow Induced Vibration (FIV). In particular, the wear is localized and concentrated in a small area of upper part of U-bend in the Central Cavity region. The region has some conditions susceptible to the FIV, which are high flow velocity, high void fraction, and long unsupported span. Even though the FIV could be occurred by many mechanisms, the main mechanism would be fluid-elastic instability, or turbulent excitation. To remedy the problem, Eggcrate Flow Distribution Plate (EFDP) was installed in the Central Cavity region or Ulchin Unit 5 and 6 steam generators, so that it reduces the flow velocity in the region to a certain level. However, the cause of the FIV and the effectiveness of the EFDP was not thoroughly studied and checked. In this study, therefore the Stability Ratio (SR), which is the ratio of the actual velocity to the critical velocity, was compared between the value before the installation of EFDP and that after. Also the possibility of fluid-elastic instability of KSNP steam generator and the effectiveness of EFDP were checked based on the ATHOS3 code calculation and the Pettigrew's experimental results. The calculated results were plotted in a fluid-elastic instability criteria-diagram (Pettigrew, 1998, Fig. 9). The plotted result showed that KSNP steam generator with EFDP had the margin of Fluid-Elastic Instability by almost 25％.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2003.05a
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• pp.670-676
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• 2003

To investigate the group plugging effect the fluid-elastic instability analysis has been performed on various column and row number of the KSNP steam generator lutes. This study compares the stability ratio of the plugged tube with that of the intact one. The information on the thermal-hydraulic data of the steam generator have been obtained by using the ATHOS3-MOD1 code with and without the thermal energy transfer at the plugged region. Both of the boundary conditions of hot-leg temperature and feedwater mass flow rate are fixed for this investigation. From the results of this study the stability ratios inside the group plugging zone are decreased slightly. At the outside of group plugging zone, however, most of the stability ratios tend to be increased.

• Nuclear industry
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• v.24 no.8 s.258
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• pp.26-37
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• 2004

• Nuclear industry
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• v.23 no.5 s.243
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• pp.63-68
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• 2003

• Nuclear industry
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• v.22 no.10 s.236
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• pp.46-50
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• 2002

• Nuclear industry
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• v.20 no.12 s.214
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• pp.46-49
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• 2000

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.13 no.6
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• pp.452-459
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• 2003

The characteristics of fluid-elastic instability and effects of turbulent excitations for the KSNP steam generator tubes were investigated numerically. The information for the thermal-hydraulic data of the steam generator has been obtained by using the ATHOS3-MOD1 code and the flow-induced vibration(FIV) analysis has been conducted by using the PIAT(program for Integrity assessment of SG tube) code. The KSNP steam generator has the concentrated plugging zone at the vicinity of the stay cylinder inside the SG. To investigate the cause of the concentrated tube plugging zone, the FIV analysis has been performed for various column and row number of the steam generator tubes. From the results of FIV analysis the stability ratio due to the fluid-elastic instability and vibrational amplitude due to the turbulent excitation in the concentrated plugged zone have a trend of larger values than those of the outer concentrated tube Plugging zone.