• Journal of the Korean Society of Safety
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• v.37 no.5
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• pp.80-88
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• 2022

The Korean nuclear industry had developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code, which adopts a two-fluid, three-field model that is comprised of gas, continuous liquid and droplet fields and has the capability to simulate three-dimensional models. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for the accident management plan of a nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification is required for the separate and integral effect experiments. Therefore, the goal of this work is to verify the calculation capability of the SPACE code for multiple failure accidents. For this purpose, an experiment was conducted to simulate a Control Element Drive Mechanism (CEDM) break with a safety injection failure using the ATLAS test facility, which is operated by Korea Atomic Energy Research Institute (KAERI). This experiment focused on the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The results of the overall system transient response using the SPACE code showed similar trends with the experimental results for parameters such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it can be concluded that the SPACE code has sufficient capability to simulate a CEDM break with a safety injection failure accident.

• Proceedings of the KIEE Conference
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• 1999.07a
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• pp.127-129
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• 1999

In this paper, we analogized effective thermal conductivity ($k_{eff}$) and effective heat capacity($pc_{p.eff}$) of insulators and the air in coil portions of linear pulse motor for CEDM(Control Element Drive Mechanism). The errors of simulated results are under 1.6% when compared with experimental results. And we carried out temperature rise of linear pulse motor for CEDM by the experiment.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.10b
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• pp.289-290
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• 2002

Stainless steel ball bearings are used in the control element drive mechanism and driving mechanisms such as step motor and gear boxes for the integral nuclear reactor, SMART. The bearings operate in pressurized pure water (primary coolant) at high temperature and should be lubricated with only this water because it is impossible to supply greases or any additional lubricant since the whole nuclear rector system should be perfectly sealed and the coolant cannot contain ingredients for bearing lubrication. Temperature of water changes from room temperature to about 120 degree Celsius and pressure rises up to 15MPa in the nuclear reactor. It can be anticipated that the frictional characteristics of the ball bearings changes according to the operating conditions, however little data are available in the literature. It is found that friction coefficient of 440C stainless steel itself does not change sharply according to temperature variation from the former research, and the friction coefficient is about 0.45 at low speed range. In this research frictional characteristics of the assembled ball bearings are investigated. A special tribometer is used to simulate the axial loading and the bearing operating conditions, temperature and pressure in the driving mechanism in the nuclear reactor. Highly purified water is used as lubricant ‘ and the water is heated up to 120 degree Celsius and pressurized to 15MPa. Friction force is monitored by the torque transducer.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.56-61
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• 2013

This paper presents the results of integrity assessment for the cracks happened in reactor vessel upper head penetration nozzles. The crack morphology for a boat sample from crack area was analyzed through microscope. The stress condition including weld residual stress around crack was analyzed using finite element analysis. From the results of crack morphology and stress condition, the crack was concluded as primary water stress corrosion cracking. The integrity of the cracked nozzle was assessed by the methodology provided in ASME Section XI. According to the assessment results, the remaining life of the cracked nozzle was 1.43 yrs. and the plant decided to repair it.

• 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.

• Computational Structural Engineering
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• v.10 no.3
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• pp.225-231
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• 1997

The dynamic characteristics of Control Element Drive Mechanism(CEDM) for Korea Standard Nuclear Power Plant are studied with the CEDM modeled as a secondary mass in a simplified two degree of freedom system, while the reactor vessel as a primary mass. The optimal .mu.-f curve is developed to reduce the response amplitudes of both primary and secondary masses. In order to improve a design it is proposed that the natural frequency ratio, f, should be converged to 0.93, the mass ratio, .mu., should not be reduced, and the result should be converged to the optimal .mu.-f curve. Optimal design for CEDM components has been carried out and the response amplitude ratios of reactor are reduced 10.5 - 19.7% while those of CEDM are reduced 6.3 - 3.4%.

• 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.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.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.65 no.11
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• pp.1848-1853
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• 2016

Control Rod Position Indicator in nuclear reactor vessel has developed for small reactor in Korea. Because of severe environment in reactor vessel, target of this study is to develop the suitable position indicator. In this study, solenoid type position indicator made of Mineral Insulated Cable(MI Cable) was introduced to adapt in severe environment. And inductance of the solenoid was used to indicate the rod position for high precision. But problem of this concept is that a linear slope of inductance is changed by temperature effect. To resolve this problem, two sensing coils were introduced for temperature compensation. A role of the sensing coil is to make reference linear equation about certain temperature. To confirm this concept, also, inductance of solenoid and sensing coils were measured at room and high temperature (${\sim}300^{\circ}C$). The results of measurement show that the position error of sensing coil between room and high temperature was about 2%. But it was identified that this error was resulted from insufficient test environment (temperature error between solenoid and sensing coils was about 2% at high temperature condition). Therefore, solenoid type position indicator shows that it is very suitable in reactor vessel as a high precision rod position indicator.

• 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.