• Proceedings of the KSME Conference
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• 2008.11a
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• pp.1132-1135
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• 2008

The sludge grown up in steam generators of nuclear power plants shortens the life-cycle of steam generators and reduces the output of power plants. So KHNP(Korea Hydro and Nuclear Power), the only nuclear power utility in Korea, removes it periodically using a steam generator lancing system during the outage of plants for an overhaul. KEPRI(Korea Electric Power Research Institute) has developed lancing systems with high pressured water nozzle for steam generators of nuclear power plants since 2001. In this paper, the design of a partial inter-tube lancing system for model F type steam generators will be described. The system is actuated without a DC motor inner steam generators because the motors in a steam generator make a trouble from high intensity of radioactivity as a break down.

• Corrosion Science and Technology
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• v.20 no.1
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• pp.22-25
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• 2021

In the past, Alloy 600 nickel-based alloys have been widely used in steam generators. However, most of them have been replaced by thermally treated alloy 690 tubes in recent years because mill annealed alloy 600 materials are known to be susceptible to stress corrosion cracking. Unlike this general perception, some steam generators using mill annealed alloy 600 tubes show excellent performance even though they are designed, manufactured, and operated in the same way. Therefore, various analyses were carried out to determine causes for the degradation of steam generators. Based on the general stress corrosion cracking mechanism, tube material susceptibility, residual stress, and sludge deposits of steam generators were compared to identify factors affecting stress corrosion cracking. It was found that mill annealed alloy 600 steam generator tubes showed higher resistance to stress corrosion cracking when the amount of sludge deposits on tube surface was smaller and residual stress generated during the fabrication was lower.

• Nuclear Engineering and Technology
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• v.33 no.1
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• pp.102-110
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• 2001

In this work, the model predictive control method was applied to a linear model and a nonlinear model of steam generators. The parameters of a linear model for steam generators are very different according to the power levels. The model predictive controller was designed for the linear steam generator model at a fixed power level. The proposed controller at the fixed power level showed good performance for any other power levels by designed changing only the input-weighting factor. As the input-weighting factor usually increases, its relative stability does so. The steam generator has some nonlinear characteristics. Therefore, the proposed algorithm has been implemented for a nonlinear model of the nuclear steam generator to verify its real performance and also, showed good performance.

• Nuclear Engineering and Technology
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• v.29 no.1
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• pp.85-91
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• 1997

Cross-flow existing in a shell-and-tube steam generator can cause a tube to vibrate. There are four regions subjected to cross-flow in Yonggwang units 3 and 4 (YGN 3 and 4) steam generators, which are of the same design as the steam generators for Palo Verde nuclear power plant Palo Verde units 1 and 2 steam generators have experienced localized oar at the comers of the cold side recirculating fluid inlet regions. A number of design modifications were made to preclude tube failure in specific regions of YGN 3 and 4 steam generators. Therefore, flow induced vibration experiments were done to determine the vibration magnitude of tubes in the economizer tube free lane region. The objective of this experiment is to demonstrate that the tube displacement is less than 0.01 inch rms at 100% of full power flow and to quantify the remaining design margin at 120ft and 140% of full power flow.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.40-47
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• 2013

The capacity factor of nuclear power plant in Korea is the highest level in the world. However, the integrity assessment of nuclear power plant is depended on foreign country. Especially, most eddy current testing systems for inspecting steam generators in nuclear power plant are currently imported from USA, Canada, and so on. Therefore, the eddy current testing system can react more active and adaptive from economic and managerial standpoint for actual nuclear power plants in Korea is required. In this paper, an eddy current testing system for inspecting steam generators in nuclear power plants is introduced. Frequency generator, analog circuit, analog digital converter circuit, and digital control circuit are composed in eddy current testing system. A benchmarking of acquisition system and acquisition software, eddynet 11i made by Zetec, and modifications are carried out based on the test environment of Korea nuclear power plants. Finally, all eddy current apparatus are integrated to inspect steam generator tubes in nuclear power plants.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.394-401
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• 2019

The failure of steam generators (SGs) due to corrosion is one of the most important problems in power plants. Impurities usually accumulate in the hot sides of SG and form deposits on the SG surfaces. In this paper, the sensitivity analysis of the accumulation of water impurities in the heat exchangers of nuclear power plants is presented. The convection-diffusion equation of the liquid phase on the heated surfaces is derived and then solved by the finite volume method. Also, the effects of the thermal-hydraulic parameters in the form of dimensionless numbers, such as $Pe_q$, $Pe_u$, $k_q$(relative solubility of impurity between the steam and water) on the impurities concentration are studied.

• Nuclear Engineering and Technology
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• v.38 no.4
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• pp.391-398
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• 2006

It is recommended to clean steam generators of nuclear power plants during plant outages. Under normal operations, sludge is created and constantly accumulates in the steam generators. The constituents of this sludge are different depending on each power plant characteristics. The sludge of the Kori Unit 1 steam generator, far example, was found to be composed of 93% ferrous oxide, 3% carbon and 1% of silica oxide and nickel oxide each. The research to develop a lancing system that would remove sludge deposits from the tubesheet of a steam generator was started in 1998 by the Korea Electric Power Research Institute (KEPRI) of the Korea Electric Power Corporation (KEPCO). The first commercial domestic lancing system in Korea, the $KALANS^(R)-I$ Lancing System, was completed in 2000 for Kori Unit 1 for cleaning the tubesheet of its Westinghouse Delta-60 steam generator. Thereafter, the success of the development and site implementation of the $KALANS^(R)-I$ lancing system for YGN Units 1&2 and Ulchin Units 3&4 was also realized in 2004 for sludge removal at those sites. The upper bundle cleaning system for Westinghouse model F steam generators is now under development.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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• pp.31-37
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• 2017

The heat transfer function or thermal performance is the most important function of the steam generator component in nuclear power plants. The declining of thermal performance, fouling does not affect the electric power of the nuclear power plant within a certain fouling level, but it affects the output when goes beyond the governor valve wide open of the turbine. The VWO steam pressure can be predicted through the thermal performance evaluation of steam generators in the nuclear power plant. In consideration of the fouling characteristics of the steam generator, methods of the thermal performance evaluation and fouling cases are reviewed, and also the critical VWO value is estimated through the actual thermal performance evaluation. It is necessary to apply the VWO theory based on the thermal performance of the steam generators.

• Proceedings of the KSME Conference
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• 2001.06e
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• pp.463-468
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• 2001

This paper presents the methodology for thermal hydraulic analysis of Pressurized Water Reactor (PWR) steam generators. Topics include porous media approach, governing equations, physical models and correlations for solid-to-fluid interaction and heat transfer and numerical solution scheme. Some details about the ATHOS3 code currently used widely for thermal hydraulic analysis of PWR steam generators in the industry are presented. The ATHOS3 code is applied to the thermal hydraulic analysis of steam generator in the Korea YGN 3&4 nuclear power plant and the computed results are presented.

• Nuclear Engineering and Technology
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• v.36 no.2
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• pp.127-141
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• 2004

Various types of integrated steam generators, which integrate IHTS and a steam generator into a single unit of equipment for an LMR, were analyzed using an analytic solution with some simplification. The analysis showed that the undesirable reversed heat transfer, of which occurrence was previously observed only in an integrated single-region bundle type, can also occur in an integrated double-region bundle type. The mechanism of the reversed heat transfer occurrence in the double-region type is explained and it is shown the mechanism in the double-region type is completely different from that in the single-region type. Based on this finding, a method for preventing the aforementioned heat transfer is suggested. The performance of the four types of the integrated steam generators is assessed. For this assessment, a SG is actually designed for each type and the optimization in the geometric parameters and flow rate are optimized.