• Journal of Energy Engineering
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• v.7 no.2
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• pp.202-208
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• 1998

As in the other industrial processes, a nuclear power plant involves a steam relieving process through which condensable steam is discharged and condensed in a subcooled pool. An analysis of steam discharge transients was carried out using the method of characteristics to determine the flow characteristics and dynamic loads of piping that are used for structural design of the piping and its supports. The analysis included not only the steam flow rate but also the flow rates of the air and water which originally exist in the pipe. The analytical model was developed for a uniform pipe with friction through which the flow was discharged into a suppression pool. Including the combinations of system elements such as reservoir, valve and branching pipe lines. The piping flow characteristics and dynamic loads were calculated by varying system pressure, pipe length, and submergence depth. It was found that the dynamic load, water clearing time and water clearing velocity at the water/air interface were dependent not only on the system pressure and temperature but also on the pipe length and submergence depth.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.5
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• pp.649-655
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• 2013

During welding, the dissimilar metal butt welds of nuclear piping are typically subjected to repair welding in order to eliminate defects that are found during post-weld inspection. It has been found that the repair weld can significantly increase the tensile residual stress in the weldment, and therefore, accurate estimation of the weld residual stress due to repair weld, especially for dissimilar metal welds using Ni-based alloy 82/182 in nuclear components, is of great importance in order to assess susceptibility to primary water stress corrosion cracking. In the present study, the stress distributions of dissimilar metal butt welds in nuclear reactor piping subjected to repair weld were investigated based on detailed nonlinear finite element analyses. Particular emphasis was placed on the variation of the stress distribution in the dissimilar metal butt weld according to the finite element welding analysis sequence for the repair welding process.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.77-87
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• 2018

In the case of high-energy line breaks in nuclear power plants, supersonic steam jet is formed due to the rapid depressurization. The steam jet can cause impingement load on the adjacent structures, piping systems and components. In order to secure the design integrity of the nuclear power plant, it is necessary to evaluate the load characteristics of the steam jet generated by high-energy pipe rupture. In the design process of nuclear power plant, jet impingement load evaluation was usually performed based on ANSI/ANS 58.2. However, U.S. NRC recently pointed out that ANSI/ANS 58.2 oversimplifies the jet behavior and that some assumptions are non-conservative. In addition, it is recommended that dynamic analysis techniques should be applied to consider transient load characteristics. Therefore, it is necessary to establish an evaluation methodology that can analyze the dynamic load characteristics of steam jet ejected when high energy pipe breaks. This research group has developed and validated the CFD analysis methodology to evaluate the transient behavior of supersonic impinging jet in the previous study. In this study, numerical study on the transient load characteristics of supersonic steam jet impingement was carried out and amplitude and frequency analysis of transient jet load was performed.

• Plant Journal
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• v.17 no.3
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• pp.37-41
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• 2021

P&ID((Piping and Instrument Diagram) is a key drawing in the engineering industry because it contains information about the units and instrumentation of the plant. Until now, simple repetitive tasks like listing symbols in P&ID drawings have been done manually, consuming lots of time and manpower. Currently, a deep learning model based on CNN(Convolutional Neural Network) is studied for drawing object detection, but the detection time is about 30 minutes and the accuracy is about 90%, indicating performance that is not sufficient to be implemented in the real word. In this study, the detection of symbols in a drawing is performed using 1-stage object detection algorithms that process both region proposal and detection. Specifically, build the training data using the image labeling tool, and show the results of recognizing the symbol in the drawing which are trained in the deep learning model.

• Journal of the Korean Society for Heat Treatment
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• v.35 no.4
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• pp.185-192
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• 2022

Variations in the microstructure and mechanical properties of API X70 steel processed by piping, electrical resistance welding (ERW), and post seam annealing (PSA) are investigated in this study. In the welding zone, some elongated pearlites are formed and grains coarsening occurs due to extra heat caused by the ERW and PSA processes. After the piping, the base metal shows continuous yielding behavior and a decrease in yield and impact strengths because mobile dislocation and back stress are introduced during the piping process. On the other hand, the ERW and PSA processes additionally decreased the impact strength of welding zone at room and low temperatures because some elongated pearlites easily act as crack initiation site and coarse ferrite grains facilitate crack propagation. As a result, the fracture surface of the welding zone specimen tested at low temperature revealed mostly cleavage fracture unlike the base metal specimen.

• Journal of the Korean Institute of Gas
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• v.21 no.6
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• pp.52-60
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• 2017

In diffusion process exhaust line during semiconductor manufacturing process, In order to improve the transportation efficiency in the piping by removing "The reaction by-product, $ZrO_2$ and The unreacted material, TEMAZ, TMA, $O_3$, etc" and "Powder being deposited", the piping temperature was raised to $80^{\circ}C$ or more by using the heater jacket, and the bellows at the rear end of the vacuum pump ruptured. So conducted a case study and try to prevent the similar accidents from occurring through case studies. The causes of the accident were analyzed as follows: the inflow of outside air due to the generation of a gap on the suction side of the vacuum pump and heating the pipe with the heater jacket resulted in the overpressure in the pipe due to the volumetric expansion of the gas generated by decomposition of the unreacted TEMAZ, It can be assumed that the most vulnerable bellows of the piping has been ruptured. In order to prevent such accidents, This study is aimed to identify the cause of pipeline rupture accident and to establish safety measures for the prevention of similar accidents by evaluating physical hazards of TEMAZ, which is assumed to be the cause of pipe rupture accident.

• Fire Science and Engineering
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• v.16 no.1
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• pp.24-29
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• 2002

The purpose of this study is developing the computer program for hydraulic design sprinkler systems have performed the means for the general use of network analysis method. The computer program is based on the theoretical concepts of the related Hazen-Williams equations, a modified Bernoulli equations, and the Hardy Cross method of pipe network analysis. Looped piping calculations are solved by using either the Hardy Cross method or the other iteration methods. While the other methods are solved using simultaneous equations, the Hardy Cross method is concerned with one loop at a time using reiterative process. Due to its simplicity the Hardy Cross method will be the primary method described in this thesis. The purpose of this study is to develope hydraulic calculation program by using algorithm for network analysis method. The development of computer program for the hydraulic design of sprinkler systems will perform the means in the performance-based sprinkler system design.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.37-43
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• 2014

Nuclear power plants(NPPs) are consisted of power production functions and safety functions preventing leakage of radiation. Operators working in NPPs shall maintain these functions during an operation period through various activities such as improvement & modification, corrective maintenance, preventive maintenance and surveillance test. According to the performance of these work activities, there are configuration changes in NPPs systems. Its changes cause the increase of safety risks(CDF) and plant trip risks. Recently, the importance of risk management is increasing gradually in the operation process of NPPs. Therefore, this paper presents the work management methods using the various risk monitoring systems during power operation and overhaul period. Also this paper suggests the optimum application ways of risk systems for work management.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.1
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• pp.8-14
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• 2018

Printed circuit heat exchangers (PCHEs) are widely used with an increasing demand for industrial applications. PCHEs are capable of operating at high temperatures and pressure. We consider a PCHE as a candidate intermediate heat exchanger type for a high temperature gas-cooled reactor (HTGR). For conventional application using stainless steels, design and manufacturing of PCHEs are well established. For applications to HTGR, knowledge of longitudinal conduction and deformation of channel is required to estimate design margin. This paper analyzes the effects of longitudinal conduction and deformation of channel on thermal performance using a code internally developed for design and analysis of PCHEs. The code has a capability of two dimensional simulations. Longitudinal conduction is estimated using the code. In HTGR operating condition, about ten percent of design margin is required to compensate thermal performance. The cross-sectional images of PCHE channels are obtained using an optical microscope. The images are processed with computer image process technique. We quantify the deformation of channel with dimensional parameters. It is found that the deformation has negative effect on structural integrity. The deformation enhances thermal performance when the shape of channel is straight in laminar flow regime. It reduces thermal performance in cases of a zigzag channel and turbulent flow regime.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.115-125
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• 2016

The prototype SFR has currently been under design by KAERI. The size of its major components is much larger than that of APR1400 and high temperature materials are applied for it. The increased size of components and those specific materials effect on material procurement, manufacturing process and fabrication facilities. The manufacturing methods are studied for Reactor Vessel/Guard Vessel, Control Rod Drive Mechanism, Heat Exchanger, Primary Pump, Reactor Vessel Internals, Steam Generator and In-Vessel Transfer Machine. The proper manufacturing methods are suggested for each component including side forging technology for ultra large forgings of Reactor Vessel to minimize the weld seams on which In-service Inspection should be conducted.