• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1677-1688
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• 2020

In the sodium-cooled fast reactor, the steam generator is a heat exchange device between sodium and water, which may cause leakage, resulting in a sodium-water reaction accident, which in turn affects the safe operation of the entire nuclear reactor. To this end, the electromagnetic vortex flowmeter is used to detect leakage of the steam generator and its signal processing method is studied in this paper. The hydraulic experiment was carried out by using water instead of liquid sodium, and the sensor output signal of the electromagnetic vortex flowmeter under different gas injection volumes was collected. The bubble noise signal is reflected by the base line of the sensor output signal. According to the relationship between the proportion of the bubble noise signal in the sensor output signal and the gas injection volume, a signal processing method based on the energy ratio calculation is proposed to detect whether the water contains bubbles. The gas injection experiment of liquid sodium was conducted to verify the effectiveness of the signal processing method in the detection of bubbles in sodium, and the minimum detectable leak rate of water in the steam generator was detected to be 0.2 g/s.

• Nuclear Engineering and Technology
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• v.23 no.3
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• pp.299-305
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• 1991

Corrosive environment may be generated within steam generators from condenser cooling water in-leakage. Theoretical analysis of the accumulation of chloride as a sea water impurity is being carried out for the condenser cooling water used at YGN-1,2 nuclear power stations. Calculations have shown that highly concentrated chloride solution would be produced within the steam generators in the case of sea water in-leakage. Maximum allowable design condenser leak rate(0.5 gpm) leads chloride concentration of 2.3 ppm at steam generetor and 0.6 ppm at hotwell with the maximum blowdown rate and condensate purification. Concentration factor at steam generator is dependent only on both blowdoum rate and condensate purification efficiency as follows, Concentration Factor（equation omitted）(B$\neq$O) Blowdown and condensate purification are evaluated as the only effective measures to remove impurities from the secondary systems.

• 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

• Explosives and Blasting
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• v.17 no.4
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• pp.8-17
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• 1999

Unfortunately leaks occur in heat exchangers periodically, usually at the tube to tubeplate joint. The usual method of repair is to plug off the defective area and isolate the tubes of concern from the circuit. If the leaks continua the thermal capacity of the units is progressively reduced and for this reason the alternative of using an internal bridging sleeve has been examined. This paper discusses the overall development activities that has been found necessary to bring this repair procedure to a successful conclusion for use on the nuclear steam generator. In this work we have investigated optimum explosives and explosive quality, explosive sleeving's thickness, the design of sheath stress relieving heat treatment pull-out load, hydraulic leakage, stress corrosion cracking properties. The results obtain are as follows : (1) The optimum explosives and explosive qualities are PETN and about 15~40 gr/ft of explosive sleeving in nuclear steam generator. (2) Explosive sleeving's thickness is 1.1~l.4mm, If groove of 0.35mm formed in sleeve outside existed, For the hydraulic leakage is go up, explosive sleeving of formed groove are applicate tube and turnplate. (3) If the stress relieving heat treatment are experiment in $750^\circ{C}$, $850^\circ{C}$, 15 minutes Pull-out strength of sleeving 1,500~2,300kg, hydraulic leakage is $250kg/cm^2$.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.2
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• pp.110-115
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• 2010

This study was conducted to estimate the feasibility using thermal image measurement that is applicable to internal leak diagnosis for the power plant valve. Abnormal heating of valve surface associated with high temperature steam f10w toward valve outlet side in the condition of low temperature is a primary indicator of leakage problems in high temperature and pressure valves. Thermal imaging enables to see the invisible thermal radiation that may portend impending damage before their condition becomes critical. When steam flow in valve outlet side in the condition of low temperature is converted into heat transmitted through the valve body due to the internal leakage in valve. The existence of abnormally increasable leakage rate in the valve will result in abnormally high levels of heat to be generated that can be quickly identified with a thermal image avoiding energy loss or damage of valve component. From the experimental results, it was suggested that the thermal image measurement could be an effective way to precisely diagnose and evaluate internal leak situation of valve.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.155-162
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• 2019

To reduce the secondary damage caused by leakage accidents in plant piping systems, a constant surveillance system is necessary. To ensure leaks are promptly addressed, the surveillance system should be able to detect not only the leak itself, but also the location of the leak. Recently, research to develop new methods has been conducted using cameras to detect leakage and to estimate the location of leakage. However, existing methods solely estimate whether a leak exists or not, or only provide two-dimensional coordinates of the leakage location. In this paper, a method using multiple cameras to detect leakage and estimate the three-dimensional coordinates of the leakage location is presented. Leakage is detected by each camera using MADI(Moving Average Differential Image) and histogram analysis. The two-dimensional leakage location is estimated using the detected leakage area. The three-dimensional leakage location is subsequently estimated based on the two-dimensional leakage location. To achieve this, the coordinates (x, z) for the leakage are calculated for a horizontal section (XZ plane) in the monitoring area. Then, the y-coordinate of leakage is calculated using a vertical section from each camera. The method proposed in this paper could accurately estimate the three-dimensional location of a leak using multiple cameras.

• Tribology and Lubricants
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• v.22 no.2
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• pp.66-72
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• 2006

Leakage reduction through annular type labyrinth seals of steam turbine is necessary for enhancing their efficiency and the precise prediction method of seal leakage is needed. In this study, numerical analysis for leakage prediction of the combination-type-staggered-labyrinth seal has been carried out using FLUENT 6 which is commercial CFD (Computational Fluid Dynamics) code based on FVM (Finite Volume Method) and SIMPLE algorism. The present CFD results are verified with the theoretical analysis based on Bulk-flow concept which has been mainly used in predicting seal leakage. Comparing with the result of Bulk-flow model analysis, the leakage result of CFD analysis shows good agreement within 7.1% error.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.6
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• pp.732-738
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• 2004

One of the main degradation mechanisms in steam generator tubes is stress corrosion cracking induced by residual stress. The resulting damages can cause tube bursting or leakage of the primary water which contains radioactivity. Shot peening technique has been used to prevent stress corrosion crack growth in steam generator tubes. In order to investigate the shot peening effect on stress corrosion cracking stress intensity factors are calculated for the semi-elliptical surface crack which is located in residual stress region. The residual stress distribution in steam generator tubes is obtained from the simple model proposed by Frederick et al.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.610-616
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• 2003

Recently, the common concern of nuclear power industry in the development of technology mitigating and preventing the aging of steam generator tubes prevails, because the trends of steam generator flaws at Uljin unit #1,2 and KSNP(Korea Standard Nuclear Power Plant) impose a burden on the operation of nuclear power plant. While the regulatory agency is demanding the establishment of the advanced general performance maintenance system, the steam generator management program adapting advanced technology is being developed which may comply with EPRI PWR SG Guidelines based on NEI 97-06 ‘ General Guidelines including all the maintenance aspects consist of the tube integrity assessment criteria, repair limit, allowable leakage level, water chemistry will be composed in order to obtain the approval of regulatory agency and be applied to Nuclear power plant early 2005. This presentation is to introduce maintenance state including SG tube degradation and main contents of advanced SG management program being developed, and futhermore update present and future plan, and estimate the alternation after the completion.

• Tribology and Lubricants
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• v.23 no.2
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• pp.43-48
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• 2007

Leakage reduction through annular type labyrinth seals of steam turbine is necessary for enhancing their efficiency. In this study, modified geometry of the original combination-type-staggered-labyrinth seal has been suggested and numerical analysis for leakage prediction has been carried out for the modified-combination-type-staggered-labyrinth seal both based on bulk-flow model and using the CFD code FLUENT. The theoretical analysis based on bulk-flow model yields leakage reduction of the modified combination type staggered labyrinth seal by about 11%. Comparing with the result of Bulk-flow model analysis, the leakage result of CFD analysis shows reasonable agreement within 9.8% error.