• Proceedings of the KWS Conference
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• 1996.10a
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• pp.19-34
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• 1996

A review was made of fatigue failures of nuclear power plant components in Japan, which were experienced in service and during periodical inspection. No case has been recently reported of a service fatigue failure of a reactor pressure vessel itself, excluding nozzle corner cracks, that occurred many years ago. But, service fatigue failures have been occasionally experienced in piping systems, pumps, and valves, on which fatigue design seems to have been inadequately applied. The causes of fatigue failures can be divided into two categories: mechanical-vibration-induced fatigue and thermal-fluctuation-induced fatigue. Vibration-induced fatigue failure occurs more frequently than is generally thought. The lesson gleaned from the present survey is a recognition that a service fatigue failure may occur due to any one or a combination of the following factors: (1) lack of communication between designers and fabrication engineers, (2) lack of knowledge about a possibility of fatigue failure and poor consideration about the effects of residual stresses, (3) lack of consideration on possible vibration in the design and fabrication stages, and (4) lack of fusion or poor penetration in a welded joint.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.10
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• pp.1115-1130
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• 2012

In nuclear power plants, the reactor pressure vessel (RPV) upper head control rod drive mechanism (CRDM) penetration nozzles are fabricated using J-groove weld geometry. Recently, the incidences of cracking in Alloy 600 CRDM nozzles and their associated welds have increased significantly. The cracking mechanism has been attributed to primary water stress corrosion cracking (PWSCC), and it has been shown to be driven by welding residual stresses and operational stresses in the weld region. The weld-induced residual stress is the main factor contributing to crack growth. Therefore, an exact estimation of the residual stress is important for ensuring reliable operation. This study presents the residual stress computation performed for an RPV CRDM penetration nozzle in Korea. Based on two and three dimensional finite element analyses, the effect of welding variables on the residual stress variation is estimated for sensitivity analysis.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.10 no.10
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• pp.2596-2601
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• 2009

Gouging is defined as the removal of weld metal and base metal from the opposite of a partially welded joint to facilitate complete joint penetration. Since the work by current method needs skillful welding and grinding, there is a limit on the increase of operation efficiency. Noise and dust from the weld gouging also deteriorate the work place and cause environmental problems. In this study, the gouging work by cutting method is proposed to overcome the defects from weld gouging such as low productivity, severe noise, dense dust, and so on. The developed cutting gouging system removes material as much as $13,565mm^3/min$, and enlarge the labor productivity as three times compared to that by weld gouging method.

• 제어로봇시스템학회:학술대회논문집
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• 2004.08a
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• pp.1116-1120
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• 2004

In the nuclear power plant, there are several cylindrical vessels such as reactor vessel, pressuriser and so on. The vessels are usually constructed by welding large rolled plates, forged sections or nozzle pipes together. In order to assure the integrity of the vessel, these welds should be periodically inspected using sensors such as ultrasonic transducer or visual cameras. This inspection is usually conducted under water to minimize exposure to the radioactively contaminated vessel walls. The inspections have been performed by using a conventional inspection machine with a big structural sturdy column, however, it is so huge and heavy that maintenance and handling of the machine are extremely difficult. It requires much effort to transport the system to the site and also requires continuous use of the utility's polar crane to move the manipulator into the building and then onto the vessel. Setup beside the vessel requires a large volume of work preparation area and several shifts to complete. In order to resolve these problems, we have developed an underwater mobile robot guided by the laser pointer, and performed a series of experiments both in the mockup and in the real reactor vessel. This paper introduces our robotic inspection system and the laser guidance of the mobile robot as well as the results of the functional test.

• 전기의세계
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• v.24 no.1
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• pp.75-85
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• 1975

In order to develop the commercial frequency (60Hz) plasma torch of small capacity for material cutting, welding and other industrial heating, the A.C plasma jet generator of non-transfered type is made domestically and the electrode configurations of plasma torch are composed of two kinds of electrodes W-C and W-Cu, combined by thermal emission and field emission electrode materials. In this paper, the characteristics of input power, thermal efficiency, electrode consumption, the flame and forms of arc voltage and arc current for A.C plasma torch are investigated in relation to such variables as arc current, argon flow and magnetic field intensity to obtain the basic design data necessary to A.C plasma jet generator. The result are as follows; (1)The input power, thermal efficiency and electrode consumption are influenced greatly by argon flow, magnetic field intensity and nozzle materials. (2)A.C arc voltage and current are non-symmetrial, involving D.C Component. Due to this current of D.C Component, transformer core is saturated and a large abnormal current flows into the primary winding coil. In order to prevent this abnormal current flow, a condenser must be connected in series to the main discharge circuit. (3)The stability and sharpness of jet flame are improved more in the torch of W-C electrode configuration than in the torch of W-Cu electrode configuration.

• Proceedings of the KWS Conference
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• 2009.11a
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• pp.100-100
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• 2009

This paper is concerned with numerical analyses of residual stresses in welds and material's susceptibility to stress corrosion cracking (SCC) for the primary piping system in nuclear power plants: Both the dissimilar metal weld (DMW) for stainless steel to low alloy steel joints and the similar metal weld (SMW) for forged stainless steel to cast stainless steel joints are considered. Thermal elasto-plastic analyses using the finite element method (FEM) are performed to predict residual stresses generated in fabrication welding and its related processes for both the DMW and SMW, including effects of quenching for cast stainless steel piping, machining of the DMW root, and grinding of the SMW root. As a result, the effect of quenching should be included in the evaluation of residual stresses in the SMW for the cast stainless steel piping. It is deemed that residual stresses in both the DMW and SMW would not affect the SCC susceptibility of the welds providing that the welding processes are completed without any weld repair on the inside wall of the joint. However, the grinding process if performed on the safe-end to piping weld, would produce a high level of residual stresses in the inner surface region and thus a stress improvement process (e.g. buffing) should be considered to reduce susceptibilities to SCC.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.3
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• pp.387-395
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• 2013

Recently, several circumferential cracks were found in the control rod drive mechanism (CRDM) nozzles of U.S. nuclear power plants. According to the accident analyses, coolant leaks were caused by primary water stress corrosion cracking (PWSCC). The tensile residual stresses caused by welding, corrosion sensitive materials, and boric acid solution cause PWSCC. Therefore, an exact estimation of the residual stress is important for reliable operation. In this study, finite element simulations were conducted to investigate the effects of the tube geometry (thickness and radius) on the residual stresses in a J-groove weld for different CRDM tube locations. Two different tube locations were considered (center-hole and steepest side hill tube), and the tube radius and thickness variables ($r_o/t$=2, 3, 4) included two different reference values ($r_o$=51.6, t=16.9mm).

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.10
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• pp.1337-1345
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• 2011

In pressurized water reactors (PWRs), the reactor pressure vessel (RPV) upper head contains numerous control rod drive mechanism (CRDM) nozzles. In the last 10 years, the incidences of cracking in alloy 600 CRDM nozzles and their associated welds has increased significantly. Several axial and circumferential cracks have been found in CRDM nozzles in European PWRs and U.S. nuclear power plants. These cracks are caused by primary water stress corrosion cracking (PWSCC) and have been shown to be driven by welding residual stresses and operational stresses in the weld region. Therefore, detailed finite-element (FE) simulations for the Korea Nuclear Reactor Pressure Vessel have been conducted in order to predict the magnitudes of the weld residual stresses in the tube materials. In particular, the weld residual stress results are compared in terms for nozzle location, geometry factor$r_o$/t, geometry of fillet, and adjacent nozzle.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• v.y2005m4
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• pp.27-30
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• 2005

Hanwha Corporation succeeded in domestic development of thruster heat shield for KOMPSAT-2 propulsion subsystem partly. Thruster heat shield, one of the main components of DTM, is designed to prevent the critical radiative heat exchange between thruster and satellite during firing. To overcome the manufacturing difficulties, an electro-forming process is preferred to classical welding process. In this case, an inner diameter of a new shield will be decreased a little due to the change of manufacturing process. The interference problem between thruster nozzle and heat shield was investigated through structural analysis by KARI. Hanwha manufactured heat shield based on the analysis results. In this paper, full development process is described for design, analysis, manufacturing of heat shield.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.32 no.6
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• pp.126-133
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• 2004

MRE-1 Dual Thruster Module(DTM), which will be used in KOMPSAT(Korea Multi-Purpose Satellite), can provide reliable and cost-effective means for attitude and maneuvering control system. Thruster heat shield, one of the main components of DTM, is designed to prevent the critical radiative heat exchange between thruster and satellite during firing. To overcome the manufacturing difficulties, a electroforming process is preferred to classical welding process. In this case, an inner diameter of a new shield will be decreased a little due to the change of manufacturing process. Therefore, the interference problem between thruster nozzle and heat shield is investigated through structural analysis and their results are described in this paper.