• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2017.05a
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• pp.5-13
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• 2017

To lunch the Korea Space Launch Vehicle-II(KSLV-II), the second launch complex will be constructed on the Naro Space Center and Kerosene Filling System (KFS) will be also installed newly. KFS of KSLV-II launch complex system is being designed based on Naro Launch Complex. But this must supply fuel to fuel tanks of the vehicle with only a supply pump because KSLV-II is a 3-stage launch vehicle unlike Naro Launch Vehicle or Test Launch Vehicle (TLV). A sudden rise of pump output pressure is recognized during fuel filling scenario selection process. This occurs because return flow can not actively deal with a lot of flow change using flow control method of orifice type. To solve this problem, it is verified that fuel can be stably supplied by installation of accumulator and an appropriate adjustment of filling mode change sequence through flow analysis of various cases.

• Proceedings of the KIEE Conference
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• 2007.07a
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• pp.782-783
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• 2007

Exhaustion of fossil fuel resources and high oil price, and furthermore environmental crisis due to emission of carbon dioxide from fossil fuel are serious problems in these days. In order to overcome these problems it is necessary to find and utilize the every energy saving measures and to make maximum utilization of renewable energy resources. The objective of this paper is to develop an instrument to verify the feasibility of Co-Generation System application in an Apartment Complex.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.297-305
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• 2018

Fuel assembly (FA) bow in pressurized water reactor (PWR) cores is considered to be a complex process with a large number of influencing mechanisms and several unknowns. Uncertainty and sensitivity analyses are a common way to assess the predictability of such complex phenomena. To perform such analyses, a structural model of a row of 15 FAs in the reactor core is implemented with the finite-element code ANSYS Mechanical APDL. The distribution of lateral hydraulic forces within the core row is estimated based on a two-dimensional Computational Fluid Dynamics model with porous media, assuming symmetric or asymmetric core inlet and outlet flow profiles. The influence of the creep rate on the bow amplitude is tested based on different creep models for guide tubes and fuel rods. Different FA initial states are considered: fresh FAs or FAs with higher burnup, which may be initially straight or exhibit an initial bow from previous cycles. The simulation results over one reactor cycle demonstrate that changes in the creep rate and the hydraulic conditions may have a considerable impact on the bow amplitudes and the bow patterns. A good knowledge of the specific creep behavior and the hydraulic conditions is therefore crucial for making reliable predictions.

• Nuclear Engineering and Technology
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• v.43 no.4
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• pp.335-342
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• 2011

As part of the Department of Energy's Fuel Cycle Research and Development Program an electrochemical technology employing molten salts is being developed for recycle of metallic fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. This technology has been deployed for treatment of used fuel from the Experimental Breeder Reactor II (EBR-II) in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory. This process is based on dry (non-aqueous) technologies that have been developed and demonstrated since the 1960s. These technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including preparation of associated high-level waste forms.

• Journal of Korean Society of Occupational and Environmental Hygiene
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• v.25 no.3
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• pp.392-404
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• 2015

Objectives: This study measured and analyzed the concentrations of crystalline silica, elemental carbon and the contribution ratio of pollutants which influence environmental and respiratory disease around the Ansim Briquette Fuel Complex in Daegu, Korea. Methods: We analyzed the crystalline silica and elemental carbon in the air according to FTIR(Fourier Transform Infrared Spectroscopy) and NIOSH(National Institute of Occupation Safety and Health) method 5040, respectively. In addition, lead stable isotopes, and carbon and nitrogen stable isotopes were analyzed using MC-ICP/MS(Multi Collector-Inductively Coupled Plasma/Mass Spectrometer), and IRMS(Isotope Ratio Mass Spectrometer), respectively. Results: The concentration of crystalline silica in the direct exposure area around the Ansim Briquette Fuel Complex was found to be $0.0014{\pm}0.0005mg/Sm^3$, but not to exceed the exposure standards of the ACGIH(American Conference of Governmental Industrial Hygienists). In the case of the autumn, the direct exposure area was found to show a level 2.5 times higher than the reference area, and on the whole, the direct exposure area was found to have a level 1.4 times higher than the reference area. The concentration of elemental carbon in the direct exposure area and in the reference area were found to be $0.0014{\pm}0.0006mg/Sm^3$, and $0.0006{\pm}0.0003mg/Sm^3$, respectively. This study confirmed the contribution ratio of coal raw materials to residentially deposited dusts in the area within 500 meters from the Ansim Briquette Fuel Complex and the surrounding area with a stable isotope ratio of 24.0%(0.7-62.7%) on average in the case of carbon and nitrogen, and 33.9%(26.6-54.1%) on average in the case of lead stable isotopes. Conclusions: This study was able to confirm correlations with coal raw materials used by the Ansim Briquette Fuel Complex and the surrounding area. The concentration of some pollutants, crystalline silica, and elemental carbon emitted to the direct-influence area around the Ansim Briquette Fuel Complex were relatively higher than in the reference area. Therefore, we need to impose continuous and substantive reduction countermeasures in the future to prevent particulate matter and coal raw materials in the study area. It is time for the local government and authorities to prepare active administrative methods such as the relocation of Ansim Briquette Fuel Complex.

• Corrosion Science and Technology
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• v.7 no.4
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• pp.208-211
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• 2008

Safety operation of main pipelines is primarily provided by anticorrosive monitoring. Anticorrosive monitoring of oil pipeline transportation objects is based on results of complex corrosion inspections, analysis of basic data including design data, definition of a corrosion residual rate and diagnostic of general equipment's technical condition. All the abovementioned arrangements are regulated by normative documents. For diagnostics of corrosion-technical condition of oil pipeline transportation objects one presently uses different methods such as in-line inspection using devices with ultrasonic, magnetic or another detector, acoustic-emission diagnostics, electrometric survey, general external corrosion diagnostics and cameral processing of obtained data. Results of a complex of diagnostics give a possibility: $\cdot$ to arrange a pipeline's sectors according to a degree of corrosion danger; $\cdot$ to check up true condition of pipeline's metal; $\cdot$ to estimate technical condition and working ability of a system of anticorrosive protection. However such a control of corrosion technical condition of a main pipeline creates the appearance of estimation of a true degree of protection of an object if values of protective potential with resistive component are taken into consideration only. So in addition to corrosive technical diagnostics one must define a true residual corrosion rate taking into account protective action of electrochemical protection and true protection of a pipeline one must at times. Realized anticorrosive monitoring enables to take a reasonable decision about further operation of objects according to objects' residual life, variation of operation parameters, repair and dismantlement of objects.

• Journal of the Korean Society of Systems Engineering
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• v.17 no.2
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• pp.82-90
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• 2021

The spent fuel pool (SFP) of a nuclear power plant functions to store the spent fuel. The spent fuel pool is designed to properly remove the decay heat generated from the spent fuel. If the cooling function is lost and proper operator action is not taken, the spent fuel in the storage pool can be damaged. Probabilistic safety assessment (PSA) is a safety evaluation method that can evaluate the risk of a large and complex system. So far, the probabilistic safety assessment of nuclear power plants has been mainly performed on the reactor. This study defined the requirements and the functional architecture for the probabilistic safety assessment of the spent fuel pool (SFP-PSA) by applying the systems engineering process. And, a systematic and efficient methodology was defined according to the architecture.

• 한국태양에너지학회:학술대회논문집
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• 2008.11a
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• pp.126-131
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• 2008

The large buildings in Korea usually use the generators to control the peak load of electronic consumption during the summer. It is necessary that these generators emit carbon dioxide, since they use gas or gasoline for their fuel. This study is to analyze the data of electronic consumption and operation of the generators at COEX, one of the representative complex building clusters in Korea, and to compare to the amount of carbon dioxide emitted per 1kWh from the domestic power plant by analogizing the frequency of using the generator during the summer and the amount of fuel consumption by the capacity of the generator and estimating the amount of carbon dioxide emitted from the generator.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.393-398
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• 2021

Current risk assessment of Spent Nuclear Fuel (SNF) transportation has the problem of the incomplete risk factors consideration and the general particle diffusion model utilization. In this paper, the accident frequency calculation and the detailed simulation of the accident consequences are coupled by the integrated risk assessment method. The "man-machine-environment" three-dimensional comprehensive risk indicator system is established and quantified to characterize the frequency of the transportation accidents. Consideration of vegetation, building and turbulence effect, the standard k-ε model is updated to simulate radioactive consequence of leakage accidents under complex terrain. The developed method is applied to assess the risk of the leakage accident in the scene of the typical domestic SNF Road Transportation (SNFRT). The critical risk factors and their impacts on the dispersion of the radionuclide are obtained.

• Transactions of the Korean Society of Automotive Engineers
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• v.3 no.6
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• pp.112-122
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• 1995

As an alternative fuel producing less exhaust emissions, natural gas is of interest for use both in SI and CI engines. The potential of natural gas fuelled dual-fuel engine is considered high enough. However, much effort has to be made so that gaseous fuel is used efficiently with simultaneous minimum use of pilot oil. Hence, a simplified three-dimensional model, using a finite volume method in cylindrical coordinates, has been developed to facilitate an understanding of the dual-fuel combustion phenomena and to predict the complex interactions between the pilot distillate and natural gas. The computer model was calibrated by comparing it with the experimental results obtained from diesel engine like combustion bomb tests. In the pre-mixed natural gas combustion, the fuel burning was highly reliant on the injection condition and subsequent burning nature of the pilot distillate.