• Journal of the Korean Society of Industry Convergence
• /
• v.7 no.4
• /
• pp.425-432
• /
• 2004

This study for Customer Satisfaction(Customer Focus) by Profit security' in the field Process improvement activity and man-power upgrade in the learning of organization activity or upgrading ability of each peoples. This thesis study on the focus of KAPEC which introduce Toyota system can apply to VM, 3jeong, Right Box and Right Position), 5S, JIT(Just In Time), KAlZEN, KANBAN System, CIM, ERP, DAS an output of Factory. For strategic changes to take place in industry 3 key important factors need to be included ; integration of tasks functions and process, decentralization of information and responsibility and finally simplification of products and product structures. These describes how CIM can be implemented using these factors. This study for (1)System Integration, (2) Help Logistic Problems, (3) Partly facilitated growth. (4) Improved production planning (5) Real-time management. (6) Fast reporting (7) Productivity. Quality. Delivery Up, Cost reduction and Autonomy management, FMS in the Plant etc.

• Corrosion Science and Technology
• /
• v.13 no.3
• /
• pp.87-94
• /
• 2014

Recently, the application of bending products has been increased since the industries such as automobile, aerospace, ships, and plants greatly need the usage of pipes. For facility fabrication, bending process is one of key technologies for pipings. Induction heat bending process is composed of bending deformation by repeated local heat and cooling. Because of local heating and compressive strain, detrimental phases may be precipitated and microstructural change can be induced. This work focused on the effect of induction heat bending process on the properties of ASME SA312 TP316 stainless steel. Evaluation was done on the base metal and the bended areas before and after heat treatment. Microstructure analysis, intergranular corrosion test including Huey test, double loop electropotentiokinetic reactivation test, oxalic acid etch test, and anodic polarization test were performed. On the base of microstructural analysis, grain boundaries in bended extrados area were zagged by bending process, but there were no precipitates in grain and grain boundary and the intergranular corrosion rate was similar to that of base metal. However, pitting potentials of bended area were lower than that of the base metal and zagged boundaries was one of the pitting initiation sites. By re-annealing treatment, grain boundary was recovered and pitting potential was similar to that of the base metal.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.31 no.2 s.257
• /
• pp.277-284
• /
• 2007

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from a reactor to generate steam to drive the turbine generator. Rupture of a steam generator tube can result in release of fission products to environment outside. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining the safety of a nuclear power plant. The steam generator tubes are supported at regular intervals by support plates and rotations of the tubes are restrained. Although it has been reported that the limit load for a circumferential crack is significantly affected by boundary condition of the tube, existing limit load solutions do not consider the restraining effect of support plate correctly. In addition, there are no limit load solutions for circumferential cracks in U-bend region with the effect of the support plate. This paper provides detailed limit load solutions for circumferential cracks in top of tube sheet and the U-bend regions of the steam generator tube with the actual boundary conditions to simulate the restraining effect of the support plate. Such solutions are developed based on three dimensional finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.

• Nuclear Engineering and Technology
• /
• v.46 no.5
• /
• pp.569-580
• /
• 2014

Seismic isolation is a viable strategy for protecting safety-related nuclear structures from the effects of moderate to severe earthquake shaking. Although seismic isolation has been deployed in nuclear structures in France and South Africa, it has not seen widespread use because of limited new build nuclear construction in the past 30 years and a lack of guidelines, codes and standards for the analysis, design and construction of isolation systems specific to nuclear structures. The funding by the United States Nuclear Regulatory Commission of a research project to the Lawrence Berkeley National Laboratory and MCEER/University at Buffalo facilitated the writing of a soon-to-be-published NUREG on seismic isolation. Funding of MCEER by the National Science Foundation led to research products that provide the technical basis for a new section in ASCE Standard 4 on the seismic isolation of safety-related nuclear facilities. The performance expectations identified in the NUREG and ASCE 4 for seismic isolation systems, and superstructures and substructures are described in the paper. Robust numerical models capable of capturing isolator behaviors under extreme loadings, which have been verified and validated following ASME protocols, and implemented in the open source code OpenSees, are introduced.

• Corrosion Science and Technology
• /
• v.18 no.5
• /
• pp.175-181
• /
• 2019

A failed spent fuel rod with 53,000 MWd/tU from a nuclear power plant was characterized, and the fission products and oxygen layer in the pellet-clad mechanical interaction region were observed using an EPMA (Electron Probe Micro-Analyzer). A sound fuel rod burned under similar conditions was used to compare and analyze, the results of the failed fuel rod. In the failed fuel rod, the oxide layer represented $10{\mu}m$ of the boundary of the cladding, and $35{\mu}m$ of the region outside the cladding. By comparison, in the sound fuel rod, the oxide layer was $8{\mu}m$, observed in the cladding boundary region. The cladding inner surface corrosion and the resulting fuel-cladding bonding were investigated using an EPMA. Zirconium existed in the bonding layer of the (U, Zr)O compound beyond the pellet cladding interaction gap of $20{\mu}m$, and composition of UZr2O3 was observed in the failed fuel rod. This paper presents the results of the EPMA examination of a spent fuel specimen, and a technique to analyze fission products in the pellet-clad mechanical interaction region.

• Nuclear Engineering and Technology
• /
• v.40 no.1
• /
• pp.69-76
• /
• 2008

Most pressurized water reactors use Venturi flow meters to measure the feedwater flow rate. However, fouling phenomena, which allow corrosion products to accumulate and increase the differential pressure across the Venturi flow meter, can result in an overestimation of the flow rate. In this study, a soft-sensing model based on fuzzy support vector regression was developed to enable accurate on-line prediction of the feedwater flow rate. The available data was divided into two groups by fuzzy c means clustering in order to reduce the training time. The data for training the soft-sensing model was selected from each data group with the aid of a subtractive clustering scheme because informative data increases the learning effect. The proposed soft-sensing model was confirmed with the real plant data of Yonggwang Nuclear Power Plant Unit 3. The root mean square error and relative maximum error of the model were quite small. Hence, this model can be used to validate and monitor existing hardware feedwater flow meters.

• Journal of Radiation Protection and Research
• /
• v.41 no.4
• /
• pp.354-358
• /
• 2016

Background: To find out the leak characteristic of research reactor 'HANARO' building in a typhoon condition Materials and Methods: MELCOR code which normally is used to simulate severe accident behavior in a nuclear power plant was used to simulate the leak rate of air and fission products from reactor hall after the shutdown of the ventilation system of HANARO reactor building. For the simulation, HANARO building was designed by MELCOR code and typhoon condition passed through Daejeon in 2012 was applied. Results and Discussion: It was found that the leak rate is $0.1%{\cdot}day^{-1}$ of air, $0.004%{\cdot}day^{-1}$ of noble gas and $3.7{\times}10^{-5}%{\cdot}day^{-1}$ of aerosol during typhoon passing. The air leak rate of $0.1%{\cdot}day^{-1}$ can be converted into $1.36m^3{\cdot}hr^{-1}$, but the design leak rate in HANARO safety analysis report was considered as $600m^3{\cdot}hr^{-1}$ under the condition of $20m{\cdot}sec^{-1}$ wind speed outside of the building by typhoon. Conclusion: Most of fission products during the maximum hypothesis accident at HANARO reactor will be contained in the reactor hall, so the direct radiation by remained fission products in the reactor hall will be the most important factor in designing emergency preparedness for HANARO reactor.

• The KSFM Journal of Fluid Machinery
• /
• v.16 no.4
• /
• pp.35-43
• /
• 2013

Most of the nuclear power plants(NPPs) adopts pressure difference type flow meters such as venturi and orifice meters for the measurement of feedwater flow rates to calculate reactor thermal power. However, corrosion products in the feedwater deposits on the flow meter by fouling as operating time goes. These effects lead to severe errors in the flow indication and then determination of reactor thermal power. The averaging BDFT, which has developed by Yun et al., has a potentiality to minimize this problem thanks to its inherent measurement principle. Therefore, it is expected that the averaging BDFT can replace the venturi meter for the feedwater pipe of steam generator of NPPs. The present work compares the amplification factor, K, based on CFD calculation against the K obtained from experiments in order to confirm whether a commercial CFD code can be applicable to the evaluation of characteristic for the averaging BDFT. In addition to this, the simulations to take into account of fouling effect are also carried out by rough wall option. The results show that the averaging BDFT is a promising flow meter for the accurate measurement of flow rates in the fouling condition of the NPPs.

• Nuclear Engineering and Technology
• /
• v.48 no.4
• /
• pp.975-981
• /
• 2016

The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

• Proceedings of the Korea Concrete Institute Conference
• /
• 2005.05b
• /
• pp.125-128
• /
• 2005

Coal ash, which is generated as a byproduct at a coal thermal power plant, can be classified into fly ash and bottom ash. Most of fly ash is recycled as an admixture for concrete, while bottom ash is not recycled but dumped into an ash landfill disposal site. So, if a technology for recycling bottom ash efficiently, which is increasingly generated year by year, is not developed, environmental problems will take place as a matter course and further an enormous economical cost will be required for construction of additional ash landfill disposal sites. In this study an optimum mix proportion design and a quality control method for utilizing the reclamation coal ash as an aggregate for secondary concrete products such as an artificial reef was successfully developed.