• 한국압력기기공학회 논문집
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• 제16권2호
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• pp.18-24
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• 2020

When a damage occurs in the nuclear fuel burning in the reactor, fission products that should be in the nuclear fuel rod are released into the reactor coolant. In this case, sipping test, a series of non-destructive inspection methods, are used to find leakage in nuclear fuel assemblies during the power plant overhaul period. In addition, the sipping test is also used to check the integrity of the spent fuel for moving to an intermediate dry storage, which is carried out as the first step of nuclear decommissioning, . In this paper, the principle and characteristics of the sipping test are described. The structure of the sipping inspection equipment is largely divided into a suction device that collects fissile material emitted from a damaged assembly and an analysis device that analyzes their nuclides. In order to make good use of the sipping technology, the radioactive level behavior of the primary system coolant and major damage mechanisms in the event of nuclear fuel damage are also introduced. This will be a reference for selecting an appropriate sipping method when dismantling a nuclear power plant in the future.

• 한국압력기기공학회 논문집
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• 제19권1호
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• pp.52-59
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• 2023

The welding part of the valve needs immediate action when leakage occurs due to cracks or damage. In order to repair leakage of the welding part, the valve must be separated from the pipe or replaced with a new valve. However, it is difficult to remove the valve while operating the power plant. This study presents a method to remove leakage by precisely processing the gap between the clamp and the incision part within 0.1mm while installed in the pipe system. If the external leakage is removed using a clamp on the welding part without removing the valve during operation, the time and cost required for maintenance can be reduced.

• 드라이브 ㆍ 컨트롤
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• 제20권4호
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• pp.35-44
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• 2023

This study aimed to analyze the performance of hydraulic systems for dental chair when long working hours makes the temperature of hydraulic fluid rise. The study was carried out in the following manner. First, 'cylinder's clearance' was reflected in the three kinds of hydraulic circuits, which were developed through the preceding study, in order to analyze oil leak. Second, 12 cases of simulations comprised of the up and down of cylinders were carried out. Third, it was determined whether the cylinder velocity of dental chair surpasses 1cm/s required in the development even in the hydraulic fluid temperature of 60℃. In conclusion, this study used SimulationX to verify the performance stability at high temperatures using three types of hydraulic circuits designed to develop a Korean unit chair.

• 한국수소및신에너지학회논문집
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• 제34권5호
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• pp.526-534
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• 2023

To evaluate the risk of leakage when using liquid hydrogen, a leakage test was conducted using liquid nitrogen in an outdoor environment rather than a laboratory environment. To assume a real-scale continuous leak, liquid nitrogen was allowed to leak for 5 minutes through a pipe with a diameter of 25.4 mm at a design spill rate of 60 L/min. The measurement system consisted of devices for climate conditions, LN2 spread and vapor clouds. The main experimental results are the liquid pool radius and the concentration of vapor cloud, and the radius of the liquid pool was compared with the numerical analysis results.

• Nuclear Engineering and Technology
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• 제52권8호
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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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• 제25권2호
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• pp.204-214
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• 1993

본 논문은 원자력발전소의 배관설계에 파단전 누설(leak-before-break : LBB) 개념이 적용됨에 따라 새롭게 해석대상이 된 분기관파단에 의한 노심지지배럴의 쉘응답을 계산한 것이다. 앞으로 직경 10인치 이상의 고에너지 배관에 대해 LBB 개념이 적용될 것으로 예상되는 바, 이 경우 LBB 적용대상에서 제외되는 유일한 1차측 배관인 3인치 가압기 분무관의 파단을 가정하였고 이때 노심 지지배럴에 가해지는 쉘응답을 구하였다. 이들 응답을 직경 10인치 이상인 배관파단시의 응답과 비교한 결과 앞으로 직경 10인치 이상의 배관에 대해 LBB 개념이 적용될 경우 배관파단에 대한 노심지지배럴의 쉘응답은 무시할 수 있음을 보였다.

• Nuclear Engineering and Technology
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• 제38권7호
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• pp.639-656
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• 2006

In order to apply the leak before break (LBB) concept to nuclear piping systems, the dynamic strain aging effect of low carbon steel materials has to be taken into account, in compliance with the requirements of the Korean Standard Review Guide (KSRG) 3.6.3-1. For this goal, J-R tests are needed for a range of various temperatures and loading rates, including dynamic loading conditions. In the dynamic loading J-R test, the unloading compliance method can not be applied to measure the crack growth and direct current potential drop (DCPD) method; this method also has a problem defining the crack initiation point. The normalization method is known as a very useful method to determine the J-R curve under dynamic loading because it does not need additional equipment or complicated loading sequences such as electric current or unloading. This method was accepted by the American Society for Testing and Materials (ASTM) as a standard test method E1820 A15 in 2001. However, it has not yet been clearly verified yet if the normalization method is sufficiently reliable to be applied to LBB. In this study, the basic background of the J-integral, LBB and dynamic loading J-R test are explained, and the current status for dynamic loading J-R test methods are reviewed from the view point of LBB for nuclear piping. In particular, the theoretical and historical background of the normalization method which has received attention recently, is summarized. Recent studies for this method are introduced and future works are suggested that may improve the reliability of LBB for nuclear piping.

• 한국가스학회지
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• 제26권1호
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• pp.20-26
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• 2022

대기 중으로 가스가 누출될 때, 누출구에서 비교적 가까운 위치에 누출 축을 따라 형성되는 인화성 증기운의 농도 및 속도에 관해 실험, 이론 및 수치해석을 이용한 결과를 비교하였다. 제트기류 확산(Jet dispersion) 과정에 대해 실험을 통해 농도를 측정하였고, 이론적 접근법으로써 자기유사성(Self-similarity) 이론을 적용하여 속도를 계산하였다. 그리고 두 가지 CFD 도구 - 서브그리드모델을 사용하는 FLACS와 범용 CFD 도구인 CFX-를 이용해 속도와 농도를 계산하여 각 결과와 비교하였다. 자기유사성 이론과 CFD의 가장 큰 차이점은 부력항에 의한 차이이며, 이는 누출원으로부터 거리가 멀어져 누출 모멘텀이 감소할 수록 영향력이 증가한다. 결과는 누출원 반경에 대한 무차원 거리와 최대 속도에 대한 무차원 속도를 이용하여 비교한다.

• 한국해양공학회지
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• 제13권3호통권33호
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• pp.36-48
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• 1999

Fatigue life and crack retardation behavior after penetration were experimentally examined using surface pre-cracked specimens of aluminium alloy 5083. The Wheeler model retardation parameter was used successfully to predict crack growth behavior after penetration. By using a crack propagation rule, the change in crack shape after penetration can be evaluated quantitatively. Advanced, waveform-based acoustic emission (AE) techniques have been successfully used to evaluate signal characteristics obtained form fatigue crack propagation and penetratin behavior in 6061 aluminum plate with surface crack under fatigue stress. Surface defects in the structural members are apt to be origins of fatigue crack growth, which may cause serious failure of the whole structure. The nondestructive analysis on the crack growth and penetration from these defects may, therefore, be one of the most important subjects on the reliability of the leak before break (LBB) design. The goal of the present study is to determine if different sources of the AE could be identified by characteristics of the waveforms produced from the crack growth and penetration. AE signals detected in four stages were found to have different signal per stage. With analysis of waveform and power spectrum in 6061 aluminum alloys with a surface crack, it is found to be capabilities on real-time monitoring for the crack propagation and penetration behavior of various damages and defects in structural members.

• Corrosion Science and Technology
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• 제12권5호
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• pp.227-232
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• 2013

A number of components installed in the secondary system of nuclear power plants are exposed to aging mechanisms such as FAC (Flow-Accelerated Corrosion), Cavitation, Flashing, and LDIE (Liquid Droplet Impingement Erosion). Those aging mechanisms can lead to thinning of the components. In April 2013, one (1) inch small bore piping branched from the main steam line experienced leakage resulting from wall thinning in a 1,000 MWe Korean PWR nuclear power plant. During the normal operation, extracted steam from the main steam line goes to condenser through the small bore piping. The leak occurred in the downstream of an orifice. A control valve with vertical flow path was placed on in front of the orifice. This paper deals with UT (Ultrasonic Test) thickness data, SEM images, and numerical simulation results in order to analyze the extent of damage and the cause of leakage in the small bore piping. As a result, it is concluded that the main cause of the small bore pipe wall thinning is liquid droplet impingement erosion. Moreover, it is observed that the leak occurred at the reattachment point of the vortex flow in the downstream side of the orifice.