• Journal of the KSME
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• v.16 no.2
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• pp.100-107
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• 1976

Stress analysis of tapered cylinder of reactor vessels is investigated by means of the intersection method. The tapered cylinder is approximated into three models-average cylinder, conical frustum, and ring. The results are compared with those of the finite element method program and an experiment. In this paper, the following results are obtained: (1) the best aproximation has been obtained by the ring model analysis: (2) the intersection analysis of the tapered cylinder by the ring model shows a sufficient accuracy for the stress analysis of reactor vessels.

• Proceedings of the KAIS Fall Conference
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• 2001.05a
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• pp.79-81
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• 2001

저압 기상 영역에서 Anhydrous HF 가스와 Methanol vapor를 사용하는 산화막 식각공정을 수행하기 위하여 (1) 반응기 부피의 최소화, (B) 공정압력의 최소화, (3) 고순도 알루미나 Reactor 적용, (4) Cluster화의 개념을 적용한 VPC 장치를 제작하였다. Wafer의 온도, HF의 분압 및 Working Pressure 등의 공정변수에 따른 Oxide Wafer의 식각특성의 변화를 확인하였다. 또한 Etch Uniformity를 향상시키기 위하여 Shower Head 구조를 변경시켜서 실험하였으며, CFD Simulation을 이용하여 Reactor내에서의 HF gas 및 Methanol vapor의 분율을 예측하였다.

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.4
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• pp.273-280
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• 2016

The simulation of nondestructive testing has been used in the prediction of the signal characteristics of various defects and in the development of the procedures. CIVA, a simulation tool dedicated to nondestructive testing, has good accuracy and speed, and provides a three-dimensional graphical user interface for improved visualization and familiar data displays consistent with an NDE technique. Even though internal validations have been performed by the CIVA software development specialists, an independent validation study is necessary for the assessment of the accuracy of the software prior to practical use. In this study, time of flight diffraction signals of ultrasonic inspection of a calibration block for reactor vessel head penetrations were simulated using CIVA. The results were compared to the experimentally inspected signals. The accuracy of the simulated signals and the possible range for simulation were verified. It was found that, there is a good agreement between the CIVA simulated and experimental results in the A-scan signal, B-scan image, and measurement of depth.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1524-1536
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• 2017

During the fabrication process of reactor vessel head penetration (RVHP), the grain size of the tube material can be changed by hot or cold work and the inner side of the tube can also be shrunk due to welding outside of the tube. Several nonregular time-of-flight diffraction (TOFD) signals were found because of deformed grains. In this paper, an investigation of nonregular TOFD indications acquired from RVHP tubes using experiments and computer simulation was performed in order to identify and distinguish TOFD signals by coarse grains from those by Primary Water Stress Corrosion Crack (PWSCC). For proper understanding of the nonregular TOFD indications, microstructural analysis of the RVHP tubes and prediction of signals scattered from the grains using Finite Element Method (FEM) simulation were performed. Prediction of ultrasonic signals from the various sizes of side drilled holes to find equivalent flaws, determination of the size of the nonregular TOFD indications from the coarse grains, and experimental investigation of TOFD signals from coarse grain and shrinkage geometry to identify PWSCC signals were performed. From the computer simulation and experimental investigation results, it was possible to obtain the nonregular TOFD indications from the coarse grains in the alloy 690 penetration tube of RVHP; these nonregular indications may be classified as PWSCC. By comparing the computer simulation and experimental results, we were able to confirm a clear difference between the coarse grain signal and the PWSCC signal.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.6
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• pp.637-647
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• 2014

In pressurized water reactors, the upper head of the reactor pressure vessel (RPV) contains numerous control rod drive mechanism (CRDM) nozzles. These nozzles are fabricated by welding after being inserted into the RPV head with a room temperature shrink fit. The tensile residual stresses caused by this welding are a major factor in primary water stress corrosion cracking (PWSCC). Over the last 15 years, the incidences of cracking in alloy 600 CRDM nozzles have increased significantly. These cracks are caused by PWSCC and have been shown to be driven by the welding residual stresses and operational stresses in the weld region. Various measures are being sought to overcome these problems. The defects resulting from the welding process are often the cause of PWSCC acceleration. Therefore, any weld defects found in the RPV manufacturing process are immediately repaired by repair welding. Detailed finite-element simulations for the Korea Nuclear Reactor Pressure Vessel were conducted in order to predict the magnitudes of the repair weld residual stresses in the tube materials.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.10
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• pp.1235-1239
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• 2012

Primary water stress corrosion cracking (PWSCC) has been observed around the weld region of control rod drive mechanism (CRDM) nozzles in nuclear power plants overseas. The weld has a J-shaped groove and it connects the CRDM nozzle with the reactor vessel upper head (RVUH). It is a dissimilar metal weld (DMW), because the CRDM is made of alloy 600 and the RVUH is made of carbon steel. In this study, finite element analysis (FEA) was performed to estimate the stress condition around the weld region. Generally, it is known that a high tensile region is more susceptible to PWSCC. FEA was performed as for the condition of welding, hydrostatic test and normal operation successively to observe how the residual stress changes due to plant condition. The FEA results show that a high tensile stress region is formed around the weld starting point on the inner surface and around the weld stop point on the outer surface.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.3086-3091
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• 2007

When a molten corium is relocated in a lower head of a reactor vessel, the ERVC (External Reactor Vessel Cooling) system is actuated as coolant is supplied into a reactor cavity to remove a decay heat from the molten corium during a severe accident. To achieve this severe accident mitigation strategy, the two-phase natural circulation flow in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. For this reason, one-dimensional natural circulation flow tests were conducted to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled-down as the half height and 1/238 rectangular channel area of the APR1400 reactor vessel. As the water inlet area increased, the natural circulation mass flow rate asymptotically increased, that is, it converged at a specific value. And the circulation mass flow rate also increased as the outlet area, injected air flow rate, and outlet height increased. But the circulation mass flow rate was not changed along with the external water level variation if the water level was higher than the outlet height.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.166-171
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• 1997

LPMS(Loose Part Monitoring System)는 원자로 및 냉각재계통내에서 발생하는 금속파편의 검출 및 분석을 위하여 사용되는 진단 장비이다. 본 논문에서는 RPV(Reactor Pressure Vessel)의 상부헤드(closure head)와 하부헤드(lower head)에서의 금속파편의 충격위치를 평가하는 LPMS를 위한 새로운 기법을 제안하고, Mock-up에서의 실험을 통하여 그 효용성을 검증하였다. 즉, 수정된 원교차법을 제안하고, 이를 반구로 모델링된 RPV의 상ㆍ하부헤드에 존재하는 금속파편의 위치평가에 적용하므로써 정확한 충격위치를 찾을 수 있음을 보였다. 이들 결과는 충격물질의 질량이나 에너지를 계산하는데 정확한 정보를 제공해 줄 수가 있다.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.29 no.3
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• pp.245-252
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• 2016

The development of Top-Mounted In-Core Instrumentation(TM-ICI) is an ongoing project to reduce the risk due to severe accidents by inserting the instrumentation into a reactor closure head instead of a reactor bottom head. As part of this project, environmental fatigue analyses for TM-ICI nozzle have been performed using two methods of NUREG/CR-6909 and Code Case N-761. TM-ICI nozzle is subjected to transient loads for level A, level B and test conditions that should be evaluated for a fatigue analysis. It is found that a cumulative usage factor considering reactor coolant environment for TM-ICI nozzle is evaluated as less than 1, which is ASME Code allowable criteria of a fatigue analysis.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.171-178
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• 2014

After the expiration of operating license of a plant, all infrastructures within the plant must be safely dismantled to the point that it no longer requires measures for radiation protection. Despite the fact that Kori 1 and Wolsong 1 are close to the expiration of their operating license, sufficient technologies for radiological characterization, decontamination and dismantling is still under development. The purpose of this study is to develop one of methods for radiological inventory assessment on measuring object by using direct measure of large component with In-Situ measurement technique. Radiological inventory was assessed by analyzing nuclide using portable gamma spectroscopy without dismantling reactor head, and the result of direct measurement was supplemented by performing indirect measurement. Radiochemical analysis were performed on surface contamination samples as well. During the study, radiological inventory of reactor vessel calculated expanding the result. Based on the result and the radioactivity variation of each radionuclides time frame for decommissioning can be decided. Thus, it is expected that during the decommissioning of plants, the result of this study will contribute to the reduction of radiation exposure to workers.