• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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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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• 제20권1호
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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 However, most of them have been replaced by thermally treated alloy 690 tubes in recent years because mill annealed alloy 600 materials are known to be susceptible to stress corrosion cracking.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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• 제29권1호
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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.

• 대한기계학회논문집A
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• 제35권9호
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• pp.1083-1090
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• 2011

소듐냉각 고속로를 개발함에 있어 최대 현안 중 하나가 증기발생기에서의 소듐-물 반응사고 가능성이다. 이를 개선하기 위해 지금까지 수십 종 이상 연구개발 되었지만 국가마다 그 사양이 다르고, 동일한 기종이 후속기에 다시 활용되지 못할 정도로 기술이 안정화 상태에 도달하지 못하였다. 최근 개발되고 있는 증기발생기의 공통적 목표는 소듐-물 반응사고의 조기감지 및 제어, 증기발생기의 검사 및 보수가 쉽게 용접개수를 줄이고 경제성을 높인 Benson 증기사이클을 적용하는 것이다. 이 논문에서는 지금까지 설계 또는 활용한 증기발생기들의 사양과 문제점을 비교, 분석하였고, 이를 토대로 현안 극복방안을 제시하였다.

• Nuclear Engineering and Technology
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• 제33권1호
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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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집E
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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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• 제38권4호
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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.

• 한국압력기기공학회 논문집
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• 제6권1호
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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.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 1997년도 춘계학술대회 논문집
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• pp.677-682
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• 1997

Vibration diagnosis of steam turbine generator is essential for safe operation. For a fast few decades, several data base systems for diagnosis of steam turbine generators have been developed and proved useful. However, there still remains a problem in using data base systems such that they require an expert engineer who has a deep insight or knowledg into the system. Moreover,such data base systems can not give any information if the input is not completely fit with data base. This paper presents an effective method for vibration diagnosis of steam turbine generators using fuzzy inference. The proposed method includes also a strategy to overcome the drawback of data base system such that one cannot obtain any information when the input is insufficient or not exact. A computer program is written to realize the entire procedure for the diagnosis. Three realistic problems are dealt with to show the effectiveness of the proposed method.

• Nuclear Engineering and Technology
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• 제36권2호
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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.