• Current Photovoltaic Research
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• v.3 no.3
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• pp.71-74
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• 2015

Soluble polyaniline was synthesized by attaching titanium isoproxide ($Ti(OPr)_4$) to the amine group of the aniline. Approximately 1 to 1 molar ratio of aniline and $Ti(OPr)_4$ was mixed and polymerized with ammonium persulfate. The FTIR result showed clear difference between $TiO_2$-aniline composite ($TiO_2An$) and $TiO_2$-polyaniline composite ($TiO_2PAn$). Although the $TiO_2An$ had negligible UV-visible absorption, the $TiO_2PAn$ showed strong absorption in the UV-visible region. Photoluminescence (PL) peaks of $TiO_2An$ were shifted toward red with the reduction of the excitation energy, which could be due to the multiple emission centers. The luminescence peak shift stopped at 501 nm. The PL spectra of $TiO_2PAn$ exhibited three emission peaks at 2.88 eV (430 nm), 2.48 eV (501 nm) and 2.22 eV (558 nm). The new emission center (2.22 eV) was observed after polymerization. Field emission scanning electron microscope image showed crack-free composite film.

• The Bulletin of The Korean Astronomical Society
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• v.36 no.2
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• pp.105.1-105.1
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• 2011

We present the modeling results of deuterium fractionation of water ice, $H_2$, and the primary deuterium isotopologues of $H3^+$ in the physical conditions associated with the star and planet formation process. We calculated the deuterium chemistry for a range of gas temperatures (Tgas~10-30 K) and ortho/para ratio (opr ) of $H_2$ based on state-to-state reaction rates and explore the resulting fractionation including the formation of a water ice mantle coating grain surfaces. We find that the deuterium fractionation exhibits the expected temperature dependence of large enrichments at low gas temperature, but only for opr-H2<0.01. More significantly the inclusion of water ice formation leads to large D/H ratios in water ice (${\geq}10^{-2}$ at 10 K) but also alters the overall deuterium chemistry. For T<20 K the implantation of deuterium into ices lowers the overall abundance of HD which reduces the efficiency of deuterium fractionation at high density. Under these conditions HD will not be the primary deuterium reservoir in the cold dense interstellar medium and $H3^+$ will be the main charge carrier in the dense centers of pre-stellar cores and the protoplanetary disk midplane.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1939-1950
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• 2019

The MELCOR code useful for a plant-specific hydrogen risk analysis has inevitable limitations in prediction of a turbulent flow of a hydrogen mixture. To investigate the accuracy of the hydrogen risk analysis by the MELCOR code, results for the turbulent gas behavior at pipe rupture accident were compared with CFX results which were verified by the American National Standard Institute (ANSI) model. The postulated accident scenario was selected to be surge line failure induced by station blackout of an Optimized Power Reactor 1000 MWe (OPR1000). When the surge line failure occurred, the flow out of the surgeline was strongly turbulent, from which the MELCOR code predicted that a substantial amount of hydrogen could be released. Nevertheless, the results indicated nonflammable mixtures owing to the high steam concentration released before the failure. On the other hand, the CFX code solving the three-dimensional fluid dynamics by incorporating the turbulence closure model predicted that the flammable area continuously existed at the jet interface even in the rising hydrogen mixtures. In conclusion, this study confirmed that the MELCOR code, which has limitations in turbulence analysis, could underestimate the existence of local combustible gas at pipe rupture accident. This clear comparison between two codes can contribute to establishing a guideline for computational hydrogen risk analysis.

• Nuclear Engineering and Technology
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• v.42 no.4
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• pp.460-467
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• 2010

In this study, the parameters of the feedwater control system (FWCS) of the OPR1000 type nuclear power plant (NPP) are optimized by response surface methodology (RSM) in order to acquire better level control performance from the FWCS. The objective of the optimization is to minimize the steam generator (SG) water level deviation from the reference level during transients. The objective functions for this optimization are relationships between the SG level deviation and the parameters of the FWCS. However, in this case of FWCS parameter optimization, the objective functions are not available in the form of analytic equations and the responses (the SG level at plant transients) to inputs (FWCS parameters) can be evaluated by computer simulations only. Classical optimization methods cannot be used because the objective function value cannot be calculated directely. Therefore, the simulation optimization methodology is used and the RSM is adopted as the simulation optimization algorithm. Objective functions are evaluated with several typical transients in NPPs using a system simulation computer code that has been utilized for the system performance analysis of actual NPPs. The results show that the optimized parameters have better SG level control performance. The degree of the SG level deviation from the reference level during transients is minimized and consequently the control performance of the FWCS is remarkably improved.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.459-465
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• 2017

The goal of this study is to evaluate the effective thermal conductivity and diffusivity of containment walls as heat sinks or passive cooling systems during nuclear power plant (NPP) accidents. Containment walls consist of steel reinforced concrete, steel liners, and tendons, and provide the main thermal resistance of the heat sinks, which varies with the volume fraction and geometric alignment of the rebar and tendons, as well as the temperature and chemical composition. The target geometry for the containment walls of this work is the standard Korean NPP OPR1000. Sample tests and numerical simulations are conducted to verify the correlations for models with different densities of concrete, volume fractions, and alignments of steel. Estimation of the effective thermal conductivity and diffusivity of the containment wall models is proposed. The Maxwell model and modified Rayleigh volume fraction model employed in the present work predict the experiment and finite volume method (FVM) results well. The effective thermal conductivity and diffusivity of the containment walls are summarized as functions of density, temperature, and the volume fraction of steel for the analysis of the NPP accidents.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2859-2865
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• 2021

One of the important issues raised from the Fukushima-Daiichi accident is the safety of multi-unit sites when simultaneous accidents occur at the site and recently a multi-unit PSA methodology is being developed worldwide. Since all operation modes of the plant should be considered in the multi-unit PSA, the accident analysis needs to be performed for shutdown operation modes, too. In this study, a station blackout during the mid-loop operation is selected as a reference scenario. The overall accident progression for the mid-loop operation is slower than that for the full-power operation because the residual heat per mass of coolant is about 6 times lower than that in the mid-loop scenario. Though the fractions of Cs released from the core to the RCS in both operation modes are almost the same, the amount of Cs delivered to the containment atmosphere is quite different due to the chemisorption in the RCS. While 45.5% of the initial inventory is chemisorbed on the RCS surfaces during the full-power operation, only 2.2% during the mid-loop operation. The containment remains intact during the mid-loop operation, though 83.9% of Cs is delivered to the containment.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3892-3901
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• 2021

The pressure drop of a moisture separator in a steam generator is the important design parameter to ensure the successful performance of a nuclear power plant. The moisture separators have a wide range of operating conditions based on the arrangement of them. The prediction of the pressure drop in a moisture separator is challenging due to the complexity of the multi-dimensional two-phase vortex flow. In this study, the moisture separator test facility using the air/water two-phase flow was used to predict the pressure drop of a moisture separator in a Korean OPR-1000 reactor. The prototypical steam/water two-phase flow conditions in a steam generator were simulated as air/water two-phase flow conditions by preserving the centrifugal force and vapor quality. A series of experiments were carried out to investigate the effect of hydraulic characteristics such as the quality and liquid mass flux on the two-phase pressure drop. A new prediction model based on the scaling law was suggested and validated experimentally using the full and half scale of separators. The suggested prediction model showed good agreement with the steam/water experimental results, and it can be extended to predict the steam/water two-phase pressure drop for moisture separators.

• The korean journal of orthodontics
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• v.49 no.4
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• pp.265-276
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• 2019

Patients with obstructive sleep apnea (OSA) whose phenotype belongs to a craniofacial vulnerability are referred from sleep doctors to orthodontists. In adults, for osseo-pharyngeal reconstruction (OPR) treatment, permanent maxillomandibular advancement (MMA) surgery and use of a temporary mandibular advancement device (MAD) are applied. This case report demonstrates successful treatment of OSA through application of phased MAD and MMA in a 16-year-old male with craniofacial deformity and residual growth potential. This patient showed skeletal and dentoalveolar changes after 7-year MAD use throughout post-adolescence, which affected the design and timing of subsequent MMA surgery, as well as post-surgical orthodontic strategy. This case report suggests that OPR treatment can be useful for treatment of OSA in post-adolescent patients, from an orthodontic point of view, in close collaboration with sleep doctors for interdisciplinary diagnosis and treatment.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1217-1233
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• 2018

Following a surge of interest in multi-unit risk in the last few years, many recent studies have suggested methods for multi-unit probabilistic safety assessment (MUPSA) and addressed several related aspects. Most of the existing studies though focused on two-unit nuclear power plant (NPP) sites or used rather simplified probabilistic safety assessment (PSA) models to demonstrate the proposed approaches. When considering an NPP site with three or more units, some approaches are inapplicable or yield very conservative results. Since the number of such sites is increasing, there is a strong need to develop and validate practical approaches to the related MUPSA. This article provides several detailed approaches that are applicable to multi-unit Level 1 PSA for sites with up to six or more reactor units. To validate the approaches, a multi-unit Level 1 PSA model is developed and the site core damage frequency is estimated for each of four representative multi-unit initiators, as well as for the case of a simultaneous occurrence of independent single-unit initiators in multiple units. For this purpose, an NPP site with six identical OPR-1000 units is considered, with full-scale Level 1 PSA models for a specific OPR-1000 plant used as the base single-unit models.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1234-1245
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• 2018

The risk of multi-unit nuclear power plants (NPPs) at a site has received considerable critical attention recently. However, current probabilistic safety assessment (PSA) procedures and computer code do not support multi-unit PSA because the traditional PSA structure is mostly used for the quantification of single-unit NPP risk. In this study, the main purpose is to develop a multi-unit Level 2 PSA method and apply it to full-power operating six-unit OPR1000. Multi-unit Level 2 PSA method consists of three steps: (1) development of single-unit Level 2 PSA; (2) extracting the mapping data from plant damage state to source term category; and (3) combining multi-unit Level 1 PSA results and mapping fractions. By applying developed multi-unit Level 2 PSA method into six-unit OPR1000, site containment failure probabilities in case of loss of ultimate heat sink, loss of off-site power, tsunami, and seismic event were quantified.