• Tribology and Lubricants
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• v.20 no.3
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• pp.140-144
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• 2004

Water-lubrication ball bearings are required to install in aqueous medium where water is used as coolant or working fluid. However water-lubricated frictional characteristics of stainless steel ball bearing is not will known compared to oil-lubricated frictional characteristics. Furthermore study on friction at high temperature is rare because bearing maintenance strategy for water-lubricated or chemicals-lubricated bearings of equipment is mostly based on change of failed bearings and parts. Ball bearings and ball screws are used to transmit power in the control rod drive mechanism for an integral reactor and are lubricated with high temperature and high pressure chemically-controlled water. Bearings and power transmitting mechanical elements for a nuclear reactor require high reliability and high performance during estimated lifetime, and their performance should be verified. In this paper, experimental research results of frictional characteristics of water-lubricated ball bearing are reported.

• 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.

• 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.

• Journal of Drive and Control
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• v.15 no.1
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• pp.10-15
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• 2018

The hydraulic cylinder seals are used not only to protect leakage of the working fluids but also to prevent incoming of foreign particles into the system. Chromium plating is generally applied to improve corrosion and wear resistance. It has been noticed that sealing surface damage occurs due to the hard foreign/wear particles contained in the hydraulic oil. In this study, a three-bodied sliding contact problem related with a PTFE seal, a spherical particle and chrome-plated steel substrate is modeled to investigate the relations to wear mechanism. Using the nonlinear finite element software, MARC/MENTAT, the deformed shapes, the von Mises and first principal stress distributions with plating thickness were compared. The sealing surface was mainly abraded by hard particles embedded in the seal. The plastic deformation of the steel substrate decreased with thicker plating. Hence it could be more effective to coat the sealing surface of a hydraulic cylinder with a hard material such as TiN, TiC and DLC.

• Nuclear Engineering and Technology
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• v.45 no.1
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• pp.73-80
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• 2013

The primary water stress corrosion cracking (PWSCC) of Alloy 600 in a PWR has been reported in the control rod drive mechanism (CRDM), pressurizer instrumentation, and the pressurizer heater sleeves. Recently, two cases of boric acid precipitation that indicated leaking of the primary cooling water were reported on the bottom head surface of steam generators (SG) in Korea. The PWSCC resistance of Ni base alloys which have intergranular carbides is higher than those which have intragranular carbides. Conversely, in oxidized acidic solutions like sodium sulfate or sodium tetrathionate solutions, the Ni base alloys with a lot of carbides at the grain boundaries and shows less stress corrosion cracking (SCC) resistance. The role of grain boundary carbides in SCC behavior of Ni base alloys was evaluated and effect of intergranular carbides on the SCC susceptibility were reviewed from the literature.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.228-235
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• 2016

Flaws at dissimilar metal welds (DMWs), such as reactor coolant systems components, Control Rod Drive Mechanism (CRDM), Bottom Mounted Instrumentation (BMI) etc., in nuclear power plants have been found. Notably, primary water stress corrosion cracking (PWSCC) in the DMWs could cause significant reliability problems at nuclear power plants. Therefore, phased array ultrasound is widely used for inspecting surface break cracks and stress corrosion cracks in DMWs. However, inspection of DMWs using phased array ultrasound has a relatively low probability of detection of cracks, because the crystalline structure of welds causes distortion and splitting of the ultrasonic beams which propagates anisotropic medium. Therefore, advanced evaluation techniques of phased array ultrasound are needed for improvement in the probability of detection of flaws in DMWs. Thus, in this study, an investigation of focusing and steering phased array ultrasound in DMWs was carried out using a time reversal technique, and an adaptive focusing technique based on finite element method (FEM) simulation. Also, evaluation of focusing performance of three different focusing techniques was performed by comparing amplitude of phased array ultrasonic signals scattered from the targeted flaw with three different time delays.

• Journal of the Korean Society for Precision Engineering
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• v.21 no.12
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• pp.174-181
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• 2004

Ultra precision positioning mechanism has widely been used on semiconductor manufacturing equipments, optical spectrum analyzer and cell manipulations. Ultra precision positioning mechanism is consisted of several actuators, sensors, guides and control systems. Its efficiency depends on each performance of components. The object of this study is to design, analysis and manufacture all of the inchworm linear motor system, which is one of the equipments embodied in ultra precision positioning mechanism. Inchworm linear motor system is consisted of a controller system and an inchworm linear motor, and its driving form is similar to a motion of spanworm. A design and manufacture of inchworm linear motor, which is consisted of three PZT actuators, a rod, two columns and a guide plate, are performed. Minimizing the von-Mises stress of the hinge using Taguchi method and simulation by FEM software optimizes the structural design in a column of flexure hinge. The designed columns and guide plates are manufactured by a W-EDM and NC-milling. A controller system, which is an apparatus to drive inchworm linear motor, can easily adjust driving conditions by varying resonance frequency and input-output voltage of actuators and amplifiers. The performance of manufactured inchworm linear motor system is verified and valuated. In the future, inchworm linear motor system will be used to make a more precision positioning by reinforcing a sensor and feedback system.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.8
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• pp.1105-1111
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• 2010

The Korea Atomic Energy Research Institutes has been developing the SMART (System integrated Modular Advanced ReacTor), an environment-friendly nuclear reactor for the generation of electricity and to perform desalination. SMART reactors can be exposed to various external and internal loads caused by seismic and coolant flows. The CRDM(control rod drive mechanism), one of structures of the SMART, is a component which is adjusting inserting amount of a control rod, controlling output of reactor power and in an emergency situation, inserting a control rod to stop the reactor. The purpose of this research is performing the analysis of dynamic characteristic to ensure safety and integrity of structure of CRDM. This paper presents two FE-models, 3-D solid model and simplified Beam model of the CRDM in the coolant, and then compared the results of the dynamic characteristic about the two FE-models using a commercial Finite Element tool, ABAQUS CAE V6.8 and ANSYS V12. Beam 4 and beam 188 of simplified-model were also compared each other. And simplified model is updated for accuracy compare to 3-D solid.

• Korean Journal of Materials Research
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• v.17 no.7
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• pp.366-370
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• 2007

Silicon nitride ($Si_3N_4$) ceramics have been considered for various components of nuclear power plants such as the mechanical seal of a reactor coolant pump (RCP), the guide roller for a control rod drive mechanism (CRDM), and a seal support, etc. Corrosion behavior of $Si_3N_4$ ceramics in a high-temperature and high-pressure water must be elucidated before they can be considered as components for nuclear power plants. In this study, the corrosion behaviors of $Si_3N_4$ ceramics containing MgO and $Al_2O_3$ as sintering aids were investigated at a hydrothermal condition ($300^{\circ}C$, 9.0 MPa) in pure water and 35 ppm LiOH solution. The corrosion reactions were controlled by a diffusion of the reactive species and/or products through the corroded layer. The grain-boundary phase was preferentially corroded in pure water whereas the $Si_3N_4$ grain seemed to be corroded at a similar rate to the grain-boundary phase in LiOH solution. Flexural strengths of the $Si_3N_4$ ceramics were significantly degraded due to the corrosion reaction. Results of this study imply that a variation of the sintering aids and/or a control (e.g., crystallization) of the grain-boundary phase are necessary to increase the corrosion resistance of $Si_3N_4$ ceramics in a high-temperature water.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.9
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• pp.1159-1168
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• 2013

In pressurized water nuclear reactors (PWRs), the reactor pressure vessel (RPV) upper head contains penetration nozzles that use a control rod drive mechanism (CRDM). The penetration nozzle uses J-groove weld geometry. Recently, the occurrence of cracking in alloy 600 CRDM penetration nozzle has increased. This is attributable to primary water stress corrosion cracking (PWSCC). PWSCC is known to be susceptible to the welding residual stress and operational stress. Generally, the tensile residual stress is the main factor contributing to crack growth. Therefore, this study investigates the effect on weld residual stress through different analysis methods for normal operating conditions using finite element analysis. In addition, this study also considers the effect of repeated normal operating condition cycles on the weld residual stress. Based on the analysis result, this paper presents a normal operating condition analysis method.