• 제어로봇시스템학회:학술대회논문집
• /
• 1997.10a
• /
• pp.848-851
• /
• 1997

A graphic simulation program is developed to assimilate the remote dismantling process of research nuclear reactors. This program makes extensive use of a commercial robot graphic instruction program. Firstly, a realistic graphic model of research reactors are built along with various dismantling equipments. Using the graphic instruction languages provided by IGRIP, then, a graphic process simulation program is developed that operates interactively with the user. Consequently, it is made possible for a process designer to visualize an arbitrary dismantling sequence and interactively modify the process. It is expected that the developed system will be utilized as an effective operator aid in both design and execution phases of remote dismantling of research reactor.

• Corrosion Science and Technology
• /
• v.7 no.2
• /
• pp.77-84
• /
• 2008

Based upon the good compatibility to neutron irradiation and high temperature water environment, austenitic stainless steels are widely used for core internal structural materials of light water reactors. But, recently, intergranular cracking was detected in the stainless steels for the core applications in some commercial PWR plants. Authors studied on the root cause of the intergranular cracking and developed the countermeasure including the alternative materials for these core applications. The intergranular cracking in these core applications are defined as an irradiation assisted mechanical cracking and irradiation assisted stress corrosion cracking. In this paper, the root cause of the intergranular cracking and its countermeasure are summarized and discussed.

• Proceedings of the KIEE Conference
• /
• 1996.07a
• /
• pp.482-485
• /
• 1996

In this paper, a predictive current control of 12-pulse parallel connected dual converter system without interphase reactors(IPR) is presented. Firstly, the characteristics of system without IPR are analyzed and compared with that of system with IPR. And the predictive current control of this system is discussed. Finally the validity of the presented system and the excellence of the predictive current control response is proved through the simulation results.

• Journal of Environmental Science International
• /
• v.3 no.4
• /
• pp.403-408
• /
• 1994

Catalytic slurry reactors, in which a solid maintained in the rom of fine particles suspended in a liquid, are frequently used in chemical and biochemical and industries. In these processes the particle loading is normally low so that the effects of particles on the liquid-film mass transfer coefacent and the gas-liquid interface area are assumed to be negligible. But it is known from the works, amongst others, that the finely powdered activated carbon can increase the gas-liquid mass transfer significantly in surface-aerated reactors. The stirred cell (13.2cm inside diameter) contained four baffles and at the stirring speeds range of 80 ∼ 300ppm, the gas-liquid interfacial area could be considered as that of the cross section of the vessel (that is, 130.1cm2). When the stirrer speeds were increased, the effective interfacial area was slightly higher than the geometric area and was obtained experimentally from the Danckwerts plots. Key Words : gas-liquid interfacial area, Duckwert's Plot stirred dell. mass transfer coefficient.

• Journal of Environmental Science International
• /
• v.12 no.7
• /
• pp.791-799
• /
• 2003

This study was carried out to get more operational characteristics of the sequencing batch biofilm reactors with media volume/reactor volume ratio of 15 %, 25 % and 35 %. Experiments were conducted to find the effects of the media packing ratio on organic matters and nutrients removal. Three laboratory scale reactors were fed with synthetic wastewater. During studies, the operation mode was fixed. The organic removal efficiency didn't show large difference among three reactor of different packing media ratios. However, from the study results, the optimum packing media ratios for biological nutrient removal was shown as 25%. The denitrifying PAOs could take up and store phosphate using nitrate as electron acceptor.

• Proceedings of the KSME Conference
• /
• 2004.11a
• /
• pp.118-123
• /
• 2004

Cast austenitic stainless steel have been widely used for primary coolant piping in light water reactors. This material is subject to thermal embrittlement at reactor operating temperature. CF-8M and CF-8A cast austenitic stainless steel is used for several components, such as primary coolant piping, elbow, pump casing, and valve bodies in light water reactors. Thermal embrittlement results in spinodal decomposition of delta-ferrite leading to decreased fracture toughness. In this study, the specimens were prepared using an accelerated aging method. The measurement of ferrite content, Charpy impact test and J-R test were performed to verify the predicting equation for aged material properties. In case of above 25% ferrite content, predicted result of J-R curve might be non-conservative.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.33 no.9
• /
• pp.643-657
• /
• 2009

This paper introduces the current nuclear experimental research activities in KAERI, the unique nuclear research institute in Korea, and the universities in Korea to solve and assess the issues which have been faced in the design of new reactors such as APR1400, SMART, GEN-IV reactors as well as fusion reactor. Also the experimental evaluations of safety for operating nuclear plants have been presented. The nuclear thermalhydraulic experiments performed in such organizations are classified the fundamental test, the separated effect test, and the integral effect test with ATLAS and SNUF. Introduction is deployed according to institutes. Finally, the future works and the direction of research voyage in the nuclear thermal-hydraulic field were suggested.

• Proceedings of the KSME Conference
• /
• 2007.05a
• /
• pp.296-299
• /
• 2007

KAERI is in the process of carrying out the Nuclear Hydrogen Development and Demonstration (NHDD) Program. The indirect cycle gas cooled reactors that produce heat at temperatures in the order of $950^{\circ}C$ are being considered in the NHDD program. For the indirect gas cooled reactors, the intermediate hear exchanger (IHX) and hot gas duct (HGD) are the main components. For the NHDD program we are in the process of establishing a conceptual design of the IHX and HGD. The pre-conceptual design activities in this study dealt with a preliminary design of the IHX and the HGD including strength and thermal expansion evaluation of the main components.

• Journal of Engineering Education Research
• /
• v.15 no.1
• /
• pp.45-52
• /
• 2012

The success history of Light Water Reactors (PWR and BWR) showed how a dominant technology could be shaped in a political and economical context. The american nuclear politics, the interest of american nuclear industry, and the accumulated technological know-hows made it possible that the not inherently safe reactor-Light Water Reactor- became a prominent reactor model. The path dependency of reactor technology on LWR kept the engineers from developing a new safer reactor, even if the severe reactor accidents occurred. In oder to increase safety of nuclear power system, we should understand the social shaping process of nuclear technology.

• Environmental Engineering Research
• /
• v.19 no.4
• /
• pp.355-361
• /
• 2014

The wastewater treatment with a two-phase expanded granular sludge bed (EGSB) system for anaerobic degradation of acetate, benzoate, terephtalate and p-toluate from purified terephtalic acid (PTA) production was studied. The feasibility and effectiveness of the system was evaluated in terms of organic oxidation by chemical oxygen demand (COD), gas production, bacterial adaptability and stability in the granular sludge. Average removal efficiencies 93.5% and 72.7% were achieved in the EGSB reactors under volumetric loading rates of $1.0-15kg-COD/m^3/day$ and terephtalate and p-toluate of 351-526 mg/L, respectively. Gas production reached total methane production rate of 0.30 L/g-COD under these conditions in the sequential EGSB reactor system. Higher strength influent COD concentration above 4.8 g-COD/L related to field conditions was fed to observe the disturbance of the EGSB reactors.