• Journal of the Korean Society of Manufacturing Technology Engineers
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• v.22 no.3_1spc
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• pp.518-524
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• 2013

Recently, probabilistic assessments of nuclear power plant components have generated interest in the nuclear industries, either for the efficient inspection and maintenance of older nuclear plants or for improving the safety and cost-effective design of newly constructed nuclear plants. In the present paper, the partial safety factor (PSF) of wall-thinned nuclear piping is evaluated based on a reliability index method, from which the effect of each statistical variable (assessment parameter) on a certain target probability is evaluated. In order to calculate the PSF of a wall-thinned pipe, a limit state function based on the load and resistance factor design (LRFD) concept is first constructed. As for the reliability assessment method, both the advanced first-order second moment (AFOSM) method and second-order reliability method (SORM) are employed to determine the PSF of each probabilistic variable. The present results can be used for developing maintenance strategies considering the priorities of input variables for structural integrity assessments of wall-thinned piping, and this PSF concept can also be applied to the optimal design of the components of newly constructed plants considering the target reliability levels.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.5
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• pp.742-749
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• 2003

Failure of a pipeline due to local wall thinning is getting more attention in the nuclear power plant industry. Although guidelines such as ANSI/ASME B31G and ASME Code Case N597 are still useful fer assessing the integrity of a wall thinned pipeline, there are some limitations in these guidelines. For instance, these guidelines consider only pressure loading and thus neglect bending loading. However, most Pipelines in nuclear power plants are subjected to internal pressure and bending moment due to dead-weight loads and seismic loads. Therefore, an assessment procedure for locally wall thinned pipeline subjected to combined loading is needed. In this paper, three-dimensional finite element(FE) analyses were performed to simulate full-scale pipe tests conducted for various shapes of wall thinned area under internal pressure and bending moment. Maximum moments based on true ultimate stress(${\alpha}$$\sub$u,t/) were obtained from FE results to predict the failure of the pipe. These results were compared with test results, which showed good agreement. Additional finite element analyses were performed to investigate the effect of key parameters, such as wall thinned depth, wall thinned angle and wall thinned length, on maximum moment. Also, the effect of internal pressure on maximum moment was investigated. Change of internal pressure did not show significant effect on the maximum moment.

• Journal of ILASS-Korea
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• v.18 no.1
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• pp.9-15
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• 2013

Wall thinning of pipeline in power plants occurs mainly by flow acceleration corrosion (FAC), cavitation erosion (C/E), liquid droplet impingement erosion (LDIE). Wall thinning by FAC and C/E has been well investigated; however, LDIE in plant industries has rarely been studied due to the experimental difficulty of setting up a long injection of highly-pressurized air. In this study, we designed a long-term experimental system for LDIE and investigate the behavior of LDIE for three kinds of materials (A106B, SS400, A6061). The main control parameter was the air-water ratio (${\alpha}$), which was defined as the volumetric ratio of water to air (0.79, 1.00, 1.72). In order to clearly understand LDIE, the spraying velocity (${\nu}$) of liquid droplets was controled larger then 160 m/s and the experiments were performed for 15 days. Therefore, this research focuses relation between erosion rate and air-water ratio on the various pipe-flow materials. NPP(nuclear power plant)'s LDIE prediction theory and management technique were drawn from the obtained data.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.617-621
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• 2003

The purpose of this paper is to analyze piping failure trend of safety pipings In domestic nuclear power plants. First, database for the piping failure was constructed with 105 data fields. The database includes plant population data, event data, and service history data. 7 kinds of piping failures in domestic NPPs were investigated. Among the 7 cases, detailed root causes were investigated for 3 cases. The first one is pipe wall thinning in main feedwater pipings of Westinghouse 3 loop type plants. The root cause of the wall thinning was flow accelerated corrosion near welding area. The next one is leak event in chemical and volume control system(CVCS) due to vibration. Some cracks occurred in socket welding area. The events showed that the integrity or socket weld is very vulnerable to vibration. The last one is also a leak event in primary sampling line in Korean standard reactor due to thermal fatigue. Although the structural integrity was not maintained by the events, there was no effect on nuclear safety in the above 3 piping failure eases.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1127-1133
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• 2021

A Spent Fuel Pool (SFP) is designed to store spent fuel assemblies in the pool. And, a SFP cooling and cleanup system cools the SFP coolant through a heat exchanger which exchanges heat with component cooling water. If the cooling system fails or interfacing pipe (e.g., suction or discharge pipe) breaks, the cooling function may be lost, probably leading to fuel damage. In order to prevent such an incident, it is required to properly cool the spent fuel assemblies in the SFP by either recovering the cooling system or injecting water into the SFP. Probabilistic safety assessment (PSA) is a good tool to assess the SFP risk when an initiating event for the SFP occurs. Since PSA has been focused on reactor-side so far, it is required to study on the framework of PSA approach for SFP and identify the key factors in terms of fuel damage frequency (FDF) through a case study. In this study, therefore, a case study of SFP-PSA on the basis of design information of APR-1400 has been conducted quantitatively, and several sensitivity analyses have been conducted to understand the impact of the key factors on FDF.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1142-1153
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• 2019

This paper presents the effect of fusion procedure on the fusion performance of the thermal butt fusion in the safety class III buried HDPE piping per various tests performed, including high speed tensile impact, free bend, blunt notched tensile, notched creep, and PENT tests. The suitability of fusion joints and qualification procedures was evaluated by comparing test results from the base material and buttfusion joints. From the notched tensile test result, it was found that the fused joints have much lower toughness than the base material. It was also identified that the notched tensile test is more desirable than the high speed tensile impact and free bend tests presented in the ASME Code Case N-755-3 as a fusion qualification test method. In addition, with regard to the single low-pressure fusion joint performances, the procedure given by the ISO 21307 was determined to be better that the one specified in the Code Case N-755-3.

• Journal of Korean Society of Water and Wastewater
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• v.26 no.3
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• pp.409-421
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• 2012

Ethanolamine (ETA) is mainly used to prevent corrosion of pipe in secondary cooling system of nuclear power plant. Condensed ETA in wastewater could increase COD and T-N when it was emitted to natural water system. Compared to conventional treatments, electrochemical oxidation process using packed bed bipolar electrodes was adopted to treat COD and T-N. According to arrangement of feeder electrode, single packed bed bipolar electrode reactor and multi-paired packed bed bipolar reactor were developed and conventional zero-valent iron (ZVI) was selected as conducting bipolar electrode. Bipolar electrodes were coordinated three-dimensionally in the reactor. The experimental results showed that COD and T-N was little removed in unit system at different pH condition (pH 8 and 11) on 100V. However, in multi-paired system that applied 600V, COD was eliminated 80.85% (anode-cathode-anode, A-C-A) and 85.11% (cathode-anode-cathode, C-A-C), respectively. T-N was also removed 96.88% (A-C-A) and 90.63% (C-A-C), simultaneously. Current efficiency was estimated both single and multi-paired system. At unit bipolar packed bed reactor, current efficiency was almost zero, however in multi-paired system, current efficiency was 300~500% at A-C-A and 250~350% at C-A-C. Current efficiency was over 100% hence it was confirmed that this system is more effective than conventional electrochemical oxidation system.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.1
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• pp.26-35
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• 2022

When piping system in a nuclear power plant is subjected to a beyond design seismic condition, it is important to accurately determine possibility of crack initiation and, if initiation occurs, its location and time. From recent experimental works on elbow pipes, it was found that the crack initiation location and crack propagation direction of the SA403 WP316 stainless steel elbow pipe were affected by the pipe thickness. In this paper, the crack initiation location and crack propagation direction for SA403 WP316 stainless steel elbow pipes with different thickness were analyzed via elastic-plastic finite element analysis. Based on FE results, the effect of the pipe thickness on different crack initiation location and crack propagation direction was analyzed using ovality, stress and strain components. It was also confirmed that the presence of internal pressure had no effect on the crack initiation location and crack propagation direction.

• Transactions of the Korean Society of Mechanical Engineers
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• v.16 no.11
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• pp.2126-2135
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• 1992

In order to prevent the decay heat removal system from failure due to air entrainment or free surface vortex in the piping system, a set of simulating experiments for the midloop operation of nuclear power plant was performed. Through these experiments, a relation between the dimensionless numbers, such as submergence H/d, froude number, reynolds number, was found. However, the effect of reynolds number was negligible for the operation conditions of Nuclear power plant. It was also found that the perturbation of the system by the disturbance such as pump start, valve operation, etc., has a strong effect on the free surface vortex. Furthermore, from a view point of reactor safery, a modified inlet device of reducer type is strongly recommendable for the prevention of air entrainment.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.5
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• pp.421-426
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• 2009

The emissivity is different because the emitted angle changes according to the position of the infrared thermography camera and object. Because of this, the temperature distribution expressed when measuring the temperature by using the infrared thermography system is not the accuracy temperature. Although the real surface temperature is constant, the temperature measured by using infrared thermography camera have error in accordance with the value of emissivity. In this paper, the temperatures of the round cylindrical object and the flat square object that heated to the equal temperature were measured by infrared thermography camera. The emissivity calibration formula and correction table are made with the affect of the view angle and emission angle form the surface temperature value. The error of measured temperature values are corrected by using the emissivity calibration formula and correction table, and apply to defect detection of the nuclear power plant pipe. From the calibration method, reliability surface temperature values were obtained.