• 한국압력기기공학회 논문집
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• 제20권1호
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• pp.39-48
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• 2024

Research reactors are operated under ambient temperature and atmospheric pressure, which is much less severe conditions compared to those in typical nuclear power plants. Due to the high temperature, heat resistant materials such as austenite stainless steel should be used for the reactors in typical nuclear power plants. Whereas, as the effect of temperature is low for research reactors, materials with high resistance to neutron irradiation, such as zircaloy and beryllium, are used. Therefore, these conditions should be considered when performing integrity assessment for research reactors. In this study, a computational technique through finite element (FE) analysis was developed considering the operating conditions and materials of research reactor when conducting integrity assessment. Neutron irradiation analysis techniques using thermal expansion analysis were proposed to consider neutron irradiation growth and swelling in zirconium alloys and beryllium. A user subroutine program that can calculate the strain rate induced by neutron irradiation creep was developed for use in the commercial analysis program Abaqus. To validate the proposed technique and the user subroutine, FE analysis results were compared with hand-calculation results, and showed good agreement. Consequently, developed technique and user subroutine are suitable for evaluating structural integrity of research reactors.

• 대한기계학회논문집A
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• 제25권5호
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• pp.823-833
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• 2001

In recent years, the subject of remaining life assessment has drawn considerable attention in chemical plants, where various structural components typically operate at high temperature an pressure. Thus a life prediction methodology accounting for high temperature creep fracture is increasingly needed for the components. Critical defects in such structures are generally found in the form of semi-elliptical surface crack, and the analysis of which is consequently an important problem in engineering fracture mechanics. On this background, we first develop an auto mesh generation program for detailed 3-D finite element analyses of axial and circumferential semi-elliptical surface cracks in a piping system. A high temperature creep fracture parameter C-integral is obtained from the finite element analyses of generated 3-D models. Post crack growth module is further appended here to calculate the amount of crack growth. Finally the remaining lives of surface cracked pipes for various analytical parameters are assessed using the developed life assessment program.

• 한국압력기기공학회 논문집
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• 제14권1호
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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.

• 한국해양공학회:학술대회논문집
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• 한국해양공학회 2004년도 학술대회지
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• pp.352-355
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• 2004

Inconel 625 is useful in variety of industrial applications because of the resistance to attack in various corrosive media at temperatures from $200^{\circ}C$ to over $1090^{\circ}C$, in combination with good law- and high temperature mechanical strength. Rencently, this material is also used widely in offshore processing piping in order to extend the maintenance tenn and improve the quality of anti-corrosion. In general, high quality weldments for this material are readily produced by commonly used processes. Not all processes are applicable to this material group, Ni-alloys. Metallurgiad characterictics or the unavailability of matching, position or suitable welding processes. Nowadays, the flux cored wire is developed and applied for the better productivity in several welding position including the vertical position. in this study, the weldability and weldment characteristics of inconel 625 are considered in FCAW weld associated with the several shielding gases($80\%Ar\;+\;20\%\;CO_2,\;50\%Ar\;+\;50\%\;CO_2,\;100\%\;CO_2$) in viewpoint of welding productivity.

• 대한기계학회논문집B
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• 제20권2호
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• pp.680-688
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• 1996

In this paper, the unsteady state calculational model is proposed for the thermal stratification analysis in the feedwater line of the PWR plant. By defining dimensionless parameters in the two-dimensional polar coordinate system and applying SIMPLE algorithm, the temperature and flow profiles due to the thermal stratification are obtained. Base on the fact that the most significant condition occurs when the fluid temperature difference between the piping ends reaches as high as 166.deg. C, the present result shows that max. Dimensionless temperature difference of 0.6 (about l00.deg. C) obtained between hot and cold sections of pipe wall at dimensionless time 47.0.

• Nuclear Engineering and Technology
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• 제43권1호
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• pp.83-88
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• 2011

To define the effect of strain rate variation from 0.04% to 0.004%/s on environmental fatigue of CF8M cast stainless steel, which is used as a primary piping material in nuclear power plants, low-cycle fatigue tests were conducted at operating pressure and temperature condition of a pressurized water reactor, 15 MPa and $315^{\circ}C$, respectively. A high-pressure and high-temperature autoclave and cylindrical solid fatigue specimens were used for the strain-controlled low-cycle environmental fatigue tests. It was observed that the fatigue life of CF8M stainless steel is shortened as the strain rate decreases. Due to the effect of test temperature, the fatigue data of NUREG-6909 appears a slightly shorter than that obtained by KEPRI at the same stress amplitude of $1{\times}10^3$ MPa. The environmental fatigue correction factor $F_{en}$'s calculated with inputs of the test data increases with high strain amplitude, while the $F_{en}$'s of NUREG-6909 remain constant regardless of strain amplitude.

• 한국안전학회지
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• 제20권3호
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• pp.9-13
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• 2005

Inconel 625 is useful in variety of industrial applications because of the resistance to attack on various corrosive media at temperature from $200^{\circ}C$ to over $1090^{\circ}C$, in combination with good low and high temperature mechanical strength. Recently, this material is also used widely in offshore processing piping in order to extend the maintenance term and improvement the quality of anti-corrosion. In general, high quality weldments for this material are readily produced by commonly used processes. Not all processes are applicable to this material group, Ni-alloys. Metallurgical characteristics or the unavailability of matching, position or suitable welding process. Nowadays, the flux cored wire is developed and applied for the better productivity in several welding position including the vertical position. In this study, the weldability and weldment characteristics(mechanical properties) of inconel 625 are considered in FCAW(Flux Cored Arc Welding) associated with the severial shielding gases($80%Ar+20%CO_2,\;50%Ar+50%CO_2,\;100%CO_2$) in viewpoint of welding productivity.

• 한국압력기기공학회 논문집
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• 제7권2호
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• pp.26-33
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• 2011

To evaluate steam oxidation behaviours of Alloy 617 and Haynes 230, oxidation test were performed at $900^{\circ}C$ in steam and $steam+20\;vol.-%\;H_2$ environments. Oxidation rate in steam condition was similar to that in air for Alloy 617, while it was slightly lower for Haynes 230. When hydrogen was added to steam, oxidation rate was enhanced. Isolated $MnTiO_3$ particle were formed on $Cr_2O_3$ oxide layer and sub layer $Cr_2O_3$ were formed in steam and $steam+20\;vol.-%\;H_2$ for Alloy 617. On the other hands, $MnCr_2O_3$ layer were formed on top of $Cr_2O_3$ oxide layer for Haynes 230. The extensive sub layer $Cr_2O_3$ formation was resulted from the oxygen inward diffusion in such environments. When hydrogen was added, the oxide morphology was changed from polygonal to platelet because of the accelerated diffusion of cations under the oxide layer. In addition, decarburized zone was extended as hydrogen participated into the reactions causing carbide dissolution.

• 한국압력기기공학회 논문집
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• 제14권1호
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• pp.24-31
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• 2018

The mechanical design procedure is studied for the PCHE(printed circuit heat exchanger) with electrochemical etched flow channels. The effective heat transfer plates of PCHE are assembled by diffusion bonding to make a module. PCHE is widely used for industrial applications due to its compactness, cost efficiency, and serviceability at high pressure and/or temperature conditions. The limitations and technical barriers of PCHE are investigated for application to nuclear components. Rules for design and fabrication of PCHE are specified in ASME Section VIII but not in ASME Section III of nuclear components. Therefore, the calculation procedure of key dimensions of PCHE is defined based on ASME section VIII. The effective heat transfer region of PCHE is defined by several key dimensions such as the flow channel radius, edge width, wall thickness, and ridge width. The mechanical design procedure of key dimensions was incorporated into a program for easy use in the PCHE design. The effect of assumptions used in the key dimension calculation on stress values is numerically investigated. A comparative analysis is done by comparing finite element analysis results for the semi-circular flow channels with the formula based sizing calculation assuming rectangular cross sections.

• 한국압력기기공학회 논문집
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• 제12권1호
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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.