• Journal of computational fluids engineering
• /
• v.2 no.1
• /
• pp.93-108
• /
• 1997

This paper describes several computational results on the various energy and environmental problems using Patankar's SIMPLE method. The specific problems included in this study are : pollutant and flammable material dispersions in open and confined areas, buoyancy-driven flow in a lake, primary clarifier for water and waste water treatment, hood ventilation in workplace. cyclone combustor and Dow chlorination reactor. A control-volume based finite-difference method is employed together with the power-law scheme. The pressure-velocity coupling is resolved by the use of the revised version of SIMPLE, says SIMPLER and SIMPLEC. The Reynolds stresses are closed using the standard or RNG κ-ε models. A nonequilibrium turbulent reaction model is developed for the application of the chlorination process in the Dow thermal reactor. Other important empirical models and physical insights appeared in this study are presented and discussed in a brief note. The computational method developed in this study is considered, in general, as a viable tool for the design and determination of the optimal operating condition of various environmental engineering system of interest.

• Journal of the Korean Society of Safety
• /
• v.31 no.6
• /
• pp.129-134
• /
• 2016

Standard design of APR+(advanced power reactor plus) was certified at 2014 by Korea regulatory body. Based on the experience gained from OPR1000 and APR1400, the APR1400 was being developed as a 1,500MWe class reactor using Korean technologies for design code, reactor coolant pump, and man-machine interface system. APR+ has been basically designed to have the seismic design basis of safe shutdown earthquake (SSE) 0.3g, a 4-train safety concept based on N+2 design philosophy, and a passive auxiliary feedwater system (PAFS). Also, safety issues on the Fukushima-type accidents have been extensively reviewed and applied to enhance APR+ safety. APR+ provides higher reliability and safety against tsunami and earthquake. The purpose of this paper is to implement probabilistic safety assessment considering these design features and to analyze sensitivity of core damage frequency for large loss of coolant accident of APR+.

• Journal of Electrical Engineering and Technology
• /
• v.9 no.1
• /
• pp.307-312
• /
• 2014

We present a full finite element analysis for plasma discharge in etching process of semiconductor circuit. The charge transport equations of hydrodynamic diffusion-drift model and the electric field equation were numerically solved in a fully coupled system by using a standard finite element procedure for transient analysis. The proposed method was applied to a real plasma reactor in order to characterize the plasma sheath that is closely related to the yield of the etching process. Throughout the plasma discharge analysis, the base electrode of reactor was tested and modified for improving the uniformity around the wafer edge. The experiment and numerical results were examined along with SEM data of etching quality. The feasibility and usefulness of the proposed method was shown by both numerical and experimental results.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.11 no.1
• /
• pp.53-60
• /
• 2015

This study was investigated to develop process technology of laser cladding with austenite stainless steel for Weld Inlay repair of dissimilar metal weld in reactor vessel in/outlet nozzles. Weld Inlay experiments were performed by laser cladding repair system consisting of common manipulator, laser apparatus and welding process scheduler, etc. Single pass welding experiments were conducted in order to obtain the optimum welding process parameters for filler wires of ER309L and Alloy 52M before multi-layer laser cladding. Based on the above obtained results, multi-layer laser cladding experiments were carried out, and welding qualities for weld specimens were estimated by PT, OM, SEM and EDS analysis. Consequently, it was revealed that multi-layer laser cladding on austenite stainless steel using filler wires of ER309L and Alloy 52M could be possible to meet ASME Code standard without any weld defect.

• Nuclear Engineering and Technology
• /
• v.47 no.3
• /
• pp.306-314
• /
• 2015

This paper examines the difference in the value of the nuclear fuel cycle cost calculated by the deterministic and probabilistic methods on the basis of an equilibrium model. Calculating using the deterministic method, the direct disposal cost and Pyro-SFR (sodium-cooled fast reactor) nuclear fuel cycle cost, including the reactor cost, were found to be 66.41 mills/kWh and 77.82 mills/kWh, respectively (1 mill = one thousand of a dollar, i.e., $10^{-3}$ $). This is because the cost of SFR is considerably expensive. Calculating again using the probabilistic method, however, the direct disposal cost and Pyro-SFR nuclear fuel cycle cost, excluding the reactor cost, were found be 7.47 mills/kWh and 6.40 mills/kWh, respectively, on the basis of the most likely value. This is because the nuclear fuel cycle cost is significantly affected by the standard deviation and the mean of the unit cost that includes uncertainty. Thus, it is judged that not only the deterministic method, but also the probabilistic method, would also be necessary to evaluate the nuclear fuel cycle cost. By analyzing the sensitivity of the unit cost in each phase of the nuclear fuel cycle, it was found that the uranium unit price is the most influential factor in determining nuclear fuel cycle costs.

• Structural Engineering and Mechanics
• /
• v.41 no.1
• /
• pp.1-24
• /
• 2012

A reinforced concrete (RC) framed structure detailed according to non-seismic detailing provisions as per Indian Standard was tested on shake table under dynamic loads. The structure had 3 main storeys and an additional storey to simulate the footing to plinth level. In plan the structure was symmetric with 2 bays in each direction. In order to optimize the information obtained from the tests, tests were planned in three different stages. In the first stage, tests were done with masonry infill panels in one direction to obtain information on the stiffness increase due to addition of infill panels. In second stage, the infills were removed and tests were conducted on the structure without and with tuned liquid dampers (TLD) on the roof of the structure to investigate the effect of TLD on seismic response of the structure. In the third stage, tests were conducted on bare frame structure under biaxial time histories with gradually increasing peak ground acceleration (PGA) till failure. The simulated earthquakes represented low, moderate and severe seismic ground motions. The effects of masonry infill panels on dynamic characteristics of the structure, effectiveness of TLD in reducing the seismic response of structure and the failure patterns of non-seismically detailed structures, are clearly brought out. Details of design and similitude are also discussed.

• Nuclear Engineering and Technology
• /
• v.41 no.9
• /
• pp.1157-1170
• /
• 2009

A variety of Generation III/III+ reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world to solve the future energy supply shortfall. Nanofluid coolants showing an improved thermal performance are being considered as a new key technology to secure nuclear safety and economics. However, it should be noted that there is a lack of comprehensible design works to apply nanofluids to Generation III+ reactor designs. In this work, the review of accident scenarios that consider expected nanofluid mechanisms is carried out to seek detailed application spots. The Axiomatic Design (AD) theory is then applied to systemize the design of nanofluid-engineered nuclear safety systems such as Emergency Core Cooling System (ECCS) and External Reactor Vessel Cooling System (ERVCS). The various couplings between Gen-III/III+ nuclear safety features and nanofluids are investigated and they try to be reduced from the perspective of the AD in terms of prevention/mitigation of severe accidents. This study contributes to the establishment of a standard communication protocol in the design of nanofluid-engineered nuclear safety systems.

• Nuclear Engineering and Technology
• /
• v.53 no.11
• /
• pp.3474-3490
• /
• 2021

Sodium-cooled fast reactor (SFR) is the preferred technology of the generation-IV fast neutron reactor, and its core body mainly uses nuclear-grade 316 stainless steel. In order to prolong the design life of SFRs to 60 years and more, it is necessary to summarize and analyze the anti-corrosion effect of nuclear grade 316 stainless steel in high temperature sodium environment. The research on sodium corrosion of nuclear grade 316 stainless steel is mainly composed of several important factors, including the microstructure of stainless steel (ferrite layer, degradation layer, etc.), the trace chemical elements of stainless steel (Cr, Ni and Mo, etc) and liquid impurity elements in sodium (O, C and N, etc), carburization and mechanical properties of stainless steel, etc. Through summarizing and constructing the sodium corrosion rate equations of nuclear grade 316 stainless steel, the stainless steel loss of thickness can be predicted. By analyzing the effects of temperature, oxygen content in sodium and velocity of sodium on corrosion rate, the basis for establishing integrity evaluation standard of SFR core components with sodium corrosion is provided.

• Journal of Korean Society of Occupational and Environmental Hygiene
• /
• v.31 no.1
• /
• pp.105-110
• /
• 2021

Objectives: The purpose of this study was to develop a new sterilization plasma device that would be proficient at sterilizing and removing pathogenic bacteria and viruses, and applicable to air purification systems. Methods: In order to understand the performance characteristics, the discharge characteristics of a packed bed DBD reactor and the ozone reduction characteristics of an ozone filter installed at the outlet of the reactor were investigated. Results: The novel packed-bed DBD reactor was proposed, and it was confirmed that the plasma discharge was uniformly and stably discharged throughout the entire layer, and sufficient ozone was generated for sterilization. The ozone filter was tested for three methods: adsorption, adsorption-decomposition, and catalytic decomposition. When the filter thickness was 30 mm, the ozone concentrations were 0.03 ppm, 0.01 ppm, and 0.21 ppm, respectively. The adsorption and adsorption-decomposition methods satisfied the EPA standard of less than 0.05ppm, but the catalytic decomposition method did not. Conclusions: It was confirmed that the adsorption-decomposition method has relatively excellent ozone filter performance and can provide the best ozone filter.

• Journal of Information Technology Services
• /
• v.20 no.3
• /
• pp.1-12
• /
• 2021

Reactor Coolant Pump (RCP) is core part of nuclear power plant to provide the forced circulation of reactor coolant for the removal of core heat. Properly monitoring vibration of RCP is a key activity of a successful predictive maintenance and can lead to a decrease in failure, optimization of machine performance, and a reduction of repair and maintenance costs. Here, we developed real-time RCP Vibration Analysis System (VAS) that web based platform using NoSQL DB (Mongo DB) to handle vibration data of RCP. In this paper, we explain how to implement digital signal process of vibration data from time domain to frequency domain using Fast Fourier transform and how to design NoSQL DB structure, how to implement web service using Java spring framework, JavaScript, High-Chart. We have implement various plot according to standard of the American Society of Mechanical Engineers (ASME) and it can show on web browser based on HTML 5. This data analysis platform shows a upgraded method to real-time analyze vibration data and easily uses without specialist. Furthermore to get better precision we have plan apply to additional machine learning technology.