• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.869-874
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• 2014

It was discovered in a Korean PWR that an extensive number of very short and shallow cracks in the SG tubes were undetectable by eddy current in-service-inspection because of the masking effect of sludge deposits. Axial stress corrosion cracks at the outside diameter of the steam generator tubes near the line contacts with the tube support plates are the major concern among the six identical Korean nuclear power plants having CE-type steam generators with Alloy 600 high temperature mill annealed tubes, HU3&4 and HB3~6. The tubes in HB3&4 have a less susceptible microstructure so that the onset of ODSCC was substantially delayed compared to HU3&4 whose tubes are most susceptible to ODSCC among the six units. The numbers of cracks detected by the eddy current inspection jumped drastically after the steam generators of HB4 were chemically cleaned. The purpose of the chemical cleaning was to mitigate stress corrosion cracking by removing the heavy sludge deposit, since a corrosive environment is formed in the occluded region under the sludge deposit. SGCC also enhances the detection capability of the eddy current inspection at the same time. Measurement of the size of each crack using the motorized rotating pancake coil probe indicated that the cracks in HB4 were shorter and substantially shallower than the cracks in HU3&4. It is believed that the cracks were shorter and shallower because the microstructure of the tubes in HB4 is less susceptible to ODSCC. It was readily understood from the size distribution of the cracks and the quantitative information available on the probability of detection that most cracks in HB4 had been undetected until the steam generators were chemically cleaned.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.92-102
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• 2017

The analysis of steam generators as an interface between primary and secondary circuits in light water nuclear power plants is crucial in terms of safety and design issues. VVER-1000 nuclear power plants use horizontal steam generators which demand a detailed thermal hydraulics investigation in order to predict their behavior during normal and transient operational conditions. Two phase flow field simulation on adjacent tube bundles is important in obtaining logical numerical results. However, the complexity of the tube bundles, due to geometry and arrangement, makes it complicated. Employment of porous media is suggested to simplify numerical modeling. This study presents the use of porous media to simulate the tube bundles within a general-purpose computational fluid dynamics code. Solved governing equations are generalized phase continuity, momentum, and energy equations. Boundary conditions, as one of the main challenges in this numerical analysis, are optimized. The model has been verified and tuned by simple two-dimensional geometry. It is shown that the obtained vapor volume fraction near the cold and hot collectors predict the experimental results more accurately than in previous studies.

• Journal of computational fluids engineering
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• v.7 no.2
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• pp.43-52
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• 2002

This paper presents the numerical methodology of ATHOS3 code for thermal hydraulic analysis of steam generators in nuclear power plant. Topics include porous media approach, governing equations, physical models and correlations for solid-to-fluid interaction and heat transfer, and numerical solution scheme. The ATHOS3 code is applied to the thermal hydraulic analysis of steam generator in the Korea Kori Unit-1 nuclear power plant and the computed results are presented

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.260-265
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• 1995

Based on a thermohydraulic estimation of the level swell and shrinkage of the nuclear steam generators, a fuzzy logic controller is designed and tested to handle the problem of controlling the level swell and its restoration. The estimation is used to form an artificial system which is nearly the opposite of the level swell and shrinkage and a PD type controller is designed to control this system. This controller is added to a PI type ordinary fuzzy logic controller to form the proposed controller which is tested through various experiments on a scaled-down steam generator. It is found to perform efficiently so that the divergence of the level to a low limit can be avoided.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.321-326
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• 1996

During the plant operation, steam generator U-tube assembly will potentially be subject to adverse environmental conditions which can cause damages to them. This report addresses the major design factors of CANDU type steam generator which are intended to minimize the potential tube damages. Such factors include U-tube material, high circulation ratio, tube-to-tubesheet joint, tube support design. Also a few suggestions are presented for the design and performance improvement of CANDU type steam generators.

• Proceedings of the IEEK Conference
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• 2002.07c
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• pp.1479-1482
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• 2002

In this work, a recursive parameter estimation algorithm estimates the mathematical model of steam generators every time step and a receding horizon controller is designed by using this estimated linear steam generator model of which parameters change as time goes on. It was shown through application to a linear model of steam generator that the proposed controller has good performance.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.220.2-220
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.551-556
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• 1997

A present study addresses a loss-of-RHR event as an initiating event under specific low power or shutdown conditions. Two typical plant configurations, cold leg opening case with water-filled steam generators and pressurizer opening case with emptied steam generators, were evaluated using the RELAP5/ MOD3.2 code. The calculation was compared with the experiment conducted at ROSA-IV/LSTF in Japan. As a result, the code was capable of simulating the system transient behavior following the event. Especially, thermal hydraulic transport processes including non-condensable gas behavior were reasonably predicted with an appropriate time step and CPU time. However, there were some code deficiencies such as too large system mass errors and severe flow oscillations in core region.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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• pp.83-88
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• 2010

Domestic nuclear steam generators with Alloy 600 HTMA tubes have experienced axial cracking at eggcrate tube support plates(TSPs). The axial stress corrosion cracks were observed at the crevice between outside of tubes and eggcrate TSPs. The root cause of axial cracking was investigated by thermal hydraulic analysis and sludge distribution diagnosis. It is suggested that deposition of sludge at eggcrate TSPs could increase the outside surface temperature of tube and promote the enrichment of impurities at crevice, and thus accelerate cracking. Additionally strategy for reducing the sludge ingress to steam generators is discussed.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.10b
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• pp.379-380
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• 2002

Steam generators (S/G) of pressurized water reactors are large heat exchangers that use the heat from the primary reactor coolant to make steam in the secondary side for driving turbine generators. Reciprocating sliding wear experiments have been performed to examine the wear properties of Incoloy 800 and Inconel 690 steam generator tubes in high temperature water. In present study, the test rig was designed to examine the reciprocating and rolling wear properties in high temperature (room temperature - $300^{\circ}C$) water. The test was performed at constant applied load and sliding distance to investigate the effect of test temperature on wear properties of steam generator tube materials. To investigate the wear mechanism of material, the worn surfaces were observed using scanning electron microscopy. At $290^{\circ}C$, wear rate of Inconel 690 was higher than that of Incoloy 800. It was assumed to be resulted from the oxide layer property difference due to the a\loy composition difference. Between 25 and $150^{\circ}C$ the wear loss increased with increasing temperature. Beyond $150^{\circ}C$, the wear loss decreased with increasing temperature. The wear loss change with temperature were due to the formation of wear protective oxide layer. From the worn surface observation, texture patterns and wear particle layers were found. As test temperature increased, the proportion of particle layer increased.