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

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.30 no.8 s.251
• /
• pp.993-1000
• /
• 2006

Operating experience of steam generators has shown that cracks of various morphology frequently occur in the steam generator tubes. These cracked tubes can stay in service if it is proved that the tubes have sufficient safety margin to preclude the risk of burst and leak. Therefore, integrity assessment using exact limit load solutions is very important for safe operation of the steam generators. This paper provides global and local limit load solutions for surface cracks in the steam generator tubes. Such solutions are developed based on three-dimensional (3-D) finite element analyses assuming elastic-perfectly plastic material behavior. For the crack location, both axial and circumferential surface cracks, and for each case, both external and internal cracks are considered. The resulting global and local limit load solutions are given in polynomial forms, and thus can be simply used in practical integrity assessment of the steam generator tubes.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2014.04a
• /
• pp.181-182
• /
• 2014

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.05a
• /
• pp.109-110
• /
• 2005

• Nuclear Engineering and Technology
• /
• v.52 no.12
• /
• pp.2880-2886
• /
• 2020

A small leak occurring on the surface of a fuel rod due to damage exposes UO₂ to a steam atmosphere. During this time, fission gas trapped inside the fuel rod leaks out, and the gas leakage can be increased due to UO₂ oxidation. Numerous studies have focused on the steam oxidation and its thermodynamic calculation in UO₂. However, the thermodynamic calculation of the UO₂ oxidation in a pressurized water reactor (PWR) environment has not been studied extensively. Moreover, the kinetics of the oxidation of UO₂ pellet also has not been investigated. Therefore, in this study, the thermodynamics of UO₂ oxidation under steam injection due to a damaged fuel rod in a PWR environment is studied. In addition, the diminishing radius of the UO₂ pellet with time in the PWR environment was calculated through an experiment simulating the initial time of steam injection at the puncture.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.16 no.1
• /
• pp.42-48
• /
• 2020

The effect of improving the tensile and J-R fracture toughness properties of SA508 Gr.1a on the LBB margin for the main steam pipe is investigated. The material properties and microstructure images of the existing main steam piping material SA106 Gr.C used in domestic nuclear power plants and the newly selected material SA508 Gr.1a were compared. For each material, LBB margins were calculated and compared through finite element analysis and crack instability evaluation. The LBB margin of the improved SA508 Gr.1a is found to be greatly improved compared to that of the existing SA106 Gr.C and SA508 Gr.1a. This is because of the increased material's strength and J-R fracture toughness compared to the previous materials. In order to analyze the effect of physical property change on the LBB margin, the sensitivity of each LBB margin according to the variation of tensile strength and J-R fracture toughness was analyzed. The effect of the change in tensile strength was found to be greater than that of the change in fracture toughness. Therefore, an increase in strength significantly influenced the improvement of the LBB margin of the improved SA508 Gr.1a.

• Journal of ILASS-Korea
• /
• v.9 no.4
• /
• pp.24-30
• /
• 2004

Feedwater flowing tube side of number 5 high pressure feedwatrr heaters was heated by extracting steam from high pressure turbine and draining water from moisture separators and number 6 high pressure feedwater heaters and supplied into steam generators. Because the extracting steam from the high pressure turbine is two phase fluid of high temperature, high pressure, and high speed and flows to inverse direction after impinging to impingement baffle. the shell wall of the number 5 high pressure feedwater heater may be affected by flow accelerated corrosion. On May 14, 1999, Point Beach Nuclear Plant (PBNP) with operating at full power experienced a steam leak from rupture of shell side of number 4B feedwater heater. Also, d domestic nuclear power plant experienced a severe wall thinning of shell side of number 5A and 5B feedwater heaters. This paper describes the fluid mixing analysis study using PHOENICS code in order to get at the root of the shell wall thinning of the feedwater heaters. The sections included in the fluid mixing analysis model are around the number 5h feedwater heater shell including the extracting pipeline. To identify the relation between the local velocities and wall thinning. the local velocities according to the analysis results were compared with the distribution of the shell wall thickness by ultrasonic test.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.34 no.9
• /
• pp.1217-1225
• /
• 2010

The possibility of a sodium-water reaction occurring in a conventional single-wall-tube steam generator in an SFR is a major problem. To improve the reliability of a steam generator, a double-wall-tube steam generator that can reduce the possibility of the occurrence of a sodium-water reaction is being developed. Current developments are focusing on improving the heat-transfer capability of a double-wall tube; further, the development of a leak-detection method to detect the occurrence of a sodium-water reaction during the reactor operation is also underway. In this study, new concepts, which will solve the above-mentioned problems, have been developed. Accordingly, a double-wall tube has been designed, fabricated, and mechanically tested for the purpose.

• Korean Chemical Engineering Research
• /
• v.57 no.1
• /
• pp.28-41
• /
• 2019

We carried out performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor. We analyzed transient-dynamic behavior of fluids inside the steam generator to vent into a sodium dump tank or a water dump tank when tubes in the steam generator were broken to cause a large-water-leak accident. Accordingly, we preliminarily evaluated design requirements of our system. Our results showed that sodium in the shell side of the steam generator and in Intermediate Heat Transport System was completely vented within 50 s and feed water in the tube side of the steam generator was completely vented within 2.5 s. It was analyzed that pressure of the tube side of the steam generator was higher than pressure of the shell side of the steam generator, which showed that sodium in the shell side did not flow into the tube side. Our results are expected to be used as basis information to performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 1992.10a
• /
• pp.81-86
• /
• 1992

원자력발전소 1차 계통 내의 건전성 감시를 위한 설비로는 음향누설 감시계 통(Acoustic Leak Monitoring System: ALMS), 금속파편 감시계통(Loose Parts Monitoring System: LPMS) 및 원자로내부구조물 진동감시계통 (Internals Vibration Monitoring System: IVMS)등이 있다. 현재, 국내의 여 러 원전에는 이들중 일부 계통들이 선택적으로 설치되어 운전중이며, 영광 3,4호기에서는 국내 최초로 이들 3개의 계통을 종합한 핵증기공급계통 건전 성감시계통(Nuclear Steam Supply System Integrity Monitoring System: NIMS)을 설계하였다. 특히, 영광 3,4호기 NIMS에서는 각 계통에 의해 감지 된 1차 계통 내의 이상상태를 하나의 분석컴퓨터(Analysis Computer)를 사 용하여 해석하는 종합결함 탐지해석 기법을 도입하였다.