• Proceedings of the KIEE Conference
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• 2002.07b
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• pp.1342-1344
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• 2002

Wind is a significant and valuable renewable energy resource. It is safe and abundant and can make an important contribution to future clean, sustainable and diversified electricity supplies. Unlike other sources of energy, wind does not pollute the atmosphere nor create any hazardous waste. In some countries wind energy is already competitive with fossil and nuclear power even without accounting for the environmental benefits of wind power. The cost of electricity from conventional power stations does not usually take full account of its environmental impact (acid rain, oil slick clean up, the effects of climate change, etc). In this paper, a transient phenomenon simulation method for Wind Power Generation System(WPGS) under real weather conditions has been proposed. The simulation method is expected to be able to analyze easily under various conditions with considering the sort of wind turbine, the capacity of system and the converter system. Wind turbine connected to the synchronous generator and power converter was simulated.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.70-71
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• 2005

It was an idea to develop a system to remove the contaminated hot particulate to make clean nuclear research facilities to the clear visibility for researcher. In order to clean micron and sub micron size particulates from used PFC solution as a decontamination tool. Since PFC solution is very expensive so it was a high recommendation to develop the some filtration equipment to purify PFC for next decontamination process, in additionally, reduce the secondary waste. However, we developed an easy and economic filter system to purify the PFC solution. The major advantage of the process operates in closed loop under near ambient conditions, thus minimizing the potential for fugitive TRU emissions and reduces the secondary waste. This has very significant safety and cost impacts. Here we proposed the two types PFC filter systems.

• Corrosion Science and Technology
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• v.16 no.3
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• pp.115-127
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• 2017

Coal tar-coated pipes buried in a domestic nuclear power plant have operated under the cathodic protection. This work conducted the simulation of the coating performance of these pipes using a FEM method. The pipes, being ductile cast iron have been suffered under considerably high cathodic protection condition beyond the appropriate condition. However, cathodic potential measured at the site revealed non-protected status. Converting from 3D CAD data of the power plant to appropriate type for a FEM simulation was conducted and cathodic potential under the applied voltage and current was calculated using primary and secondary current distribution and physical conditions. FEM simulation for coal tar-coated pipe without defects revealed over-protection condition if the pipes were well-coated. However, the simulation for coal tar-coated pipes with many defects predict that the coated pipes may be severely degraded. Therefore, for high risk pipes, direct examination and repair or renewal of pipes are strongly recommended.

• Proceedings of the SAREK Conference
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• 2008.06a
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• pp.1267-1272
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• 2008

OTEC power plants are studied as a viable option for the supply of clean energy. In this paper, the thermodynamic performance of OTEC system was calculated. The results show that the working fluids such as R32 and R125 would be alternatives based upon cutting down the system size, environmental preservation, and conditions without having a severe penalty in efficiency. the initial cost significantly. The regeneration system increase in energy efficiency, and the system can generate electricity when the difference in warm and cold seawater inlet temperatures are greater than $15^{\circ}C$. Also, the system efficiency of OTEC power plant using the condenser effluent from nuclear power plant instead of the surface water increased about 2%.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1423-1432
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• 2016

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

• Clean Technology
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• v.14 no.2
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• pp.110-116
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• 2008

The radioactive wastes generated from the nuclear industry can be divided into the forms of solid, liquid, or gas. Radioactive methyl iodide, a gaseous radioactive waste, is absorbed by activated carbon with 5 wt% of Trietylenediamine (1,4-diazania-bicycle[2.2.2]octane, TEDA) impregnated on the surface. Methyl Iodide ($CH_3I$) is combined chemically with TEDA (the final product : I-TEDA). To recycle radioactive activated carbon, removal of I-TEDA from activated carbon is needed. A wet method for recycling impregnated active carbon was developed to remove radioactive I-TEDA using an acetonitrile solution, which produces lots of secondary wastes. We suggest the removal of I-TEDA by supercritical carbon dioxide with co-solvents. In this experiment, we used a quartz crystal microbalance (QCM) for measuring the removal rate of the I-TEDA. From the experimental results, methanol was found to be the optimum co-solvent, and the optimum conditions such as temperature, pressure, and co-solvent flow rate were obtained. Possibility of using supercritical fluid in the removal of I-TEDA from radioactive activated carbon was also discussed.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.191-196
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• 1995

The intelligent human-machine interface (HMI) has been developed to enhance the safety and availability of a nuclear power plant by improving operational reliability The key elements of the HMI are the large display panels which present synopsis of the plant status and the compact, digital work stations for the primary operator control and monitoring functions. The work station consists of four consoles such as a dynamic alarm console (DAC), a system information console (SIC), a computerized operating-procedure console (COC), and a safety related information console (SRIC). The DAC provides clean alarm pictures, in which information overlapping is excluded and alarm impacts are discriminated, for quick situation awareness. The SIC covers a normal operation by offering all necessary plant information and control functions. In addition, it is closely linked with the DAC and the COC to automatically display related system information under the request of these consoles. The COC aids the operator with proper emergency operation guidelines so as to shutdown the plant safely, and it also reduces his physical/mental burden by automating the operating procedures. The SRIC continuously displays safety related information to allow the operator to assess the plant status focusing on plant safety. The proposed HMI has been validated and demonstrated with on-line data obtained from the full-scope simulator for Yonggwang Units 1,2.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.248-257
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• 2020

This study presents an initial design for a novel system consisting in a coupled nuclear reactor and a phase change material-based thermal energy storage (TES) component, which acts as a buffer and regulator of heat transfer between the primary and secondary loops. The goal of this concept is to enhance the capacity factor of nuclear power plants (NPPs) in the case of high integration of renewable energy sources into the electric grid. Hence, this system could support in elevating the economics of NPPs in current competitive markets, especially with subsidized solar and wind energy sources, and relatively low oil and gas prices. Furthermore, utilizing a prismatic-core advanced high temperature reactor (PAHTR) cooled by a molten salt with a high melting point, have the potential in increasing the system efficiency due to its high operating temperature, and providing the baseline requirements for coupling other process heat applications. The present research studies the neutronics and thermal hydraulics (TH) of the PAHTR as well as TH calculations for the TES which consists of 300 blocks with a total heat storage capacity of 150 MWd. SERPENT Monte Carlo and MCNP5 codes carried out the neutronics analysis of the PAHTR which is sized to have a 5-year refueling cycle and rated power of 300 MW_th. The PAHTR has 10 metric tons of heavy metal with 19.75 wt% enriched UO₂ TRISO fuel, a hot clean excess reactivity and shutdown margin of $33.70 and -$115.68; respectively, negative temperature feedback coefficients, and an axial flux peaking factor of 1.68. Star-CCM + code predicted the correct convective heat transfer coefficient variations for both the reactor and the storage. TH analysis results show that the flow in the primary loop (in the reactor and TES) remains in the developing mixed convection regime while it reaches a fully developed flow in the secondary loop.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1120-1130
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• 2020

The CREOLE experiment performed In the EOLE critical facility located In the Nuclear Center of CADARACHE - CEA have allowed us to get interesting and complete experimental information on the temperature effects in the light water reactor lattices. To analyze these experiments with accuracy an elaborate calculation scheme using the Monte Carlo method implemented in the MCNP6.1 code and the ENDF/B-VII.1 cross section library has been developed. We have used the ENDF/B-VII.1 data provided with the MCNP6.1.1 version in ACE format and the Makxsf utility to handle the data in the specific temperatures not available in the MCNP6.1.1 original library. The main purpose of this analysis is the qualification of the ENDF/B-VII.1 nuclear data for the prediction of the Reactivity Temperature Coefficient while ensuring the ability of the MCNP6.1 system to model such a complex experiment as CREOLE. We have analyzed the case of UO₂ lattice with 1166 ppm of boron in ordinary water moderator in specified temperatures. A detailed comparison of the calculated effective multiplication factors with the reference ones [1] in room temperature presented in this work shows a good agreement demonstrating the validation of our 3D calculation model. The discrepancies between calculations and the differential measurements of the Reactivity Temperature Coefficient for the analyzed configuration are relatively small: the maximum discrepancy doesn't exceed 1,1 pcm/℃. In addition to the analysis of direct differential measurements of the reactivity temperature coefficient performed in the poisoned UO₂ lattice configuration, we have also analyzed integral measurements in UO₂ clean lattice configuration using equivalency of the integral temperature reactivity worth with the driver core fuel reactivity worth and soluble boron reactivity worth. In this case both of the ENDF/B-VII.1 and JENDL.4 libraries were used in our analysis and the obtained results are very similar.

• Clean Technology
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• v.10 no.4
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• pp.221-228
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• 2004

Ethoxylated Nonyl Phenol Series (NP-series), nonionic surfactants, were applied for forming microemulsions in supercritical $CO_2$. Measurement results of the solubility in supercritical $CO_2$ are in the following; NP-series were high soluble in carbon dioxide in spite of the fact that those were not $CO_2$-philic surfactants traditionally well known. Water in $CO_2$ microemulsions were also formed stably. A complexation of hydrophilic lengths for $CO_2$-philic parts of NP-Series surfactants was optimized by NP-4 surfactant(N=4) for forming the microemulsions through the experiments. Formation of microemulsions was confirmed by measuring the UV-Visible spectrum through a spectroscopic method and existence of water in the microemulsions was confirmed as well. In order to apply it for a metal surface treatment or electroplating, an experiment for forming acid(organic, inorganic) solution in $CO_2$ microemulsions was carried out. Ionic surfactant in the reaction to an acid solution became unstable to form microemulsions, however, nonionic surfactant was formed stably in the reaction. Results of the study will be utilized for expanding the application scope of supercritical $CO_2$ which is an environmental-friendly solvent.