• Proceedings of the KSME Conference
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• 2003.04a
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• pp.1199-1204
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• 2003

The 3D simulations of the shielding concrete and the Rotary Specimen Rack(RSR) in the Korea Research Reactor-1&2(KRR-1&2) were carried out in present work. Four main dismantling processes, which are the removal of the RSR, reactor core region, beam tube, and thermal column and activated concrete, were selected for the graphic simulation by the consideration of the activation, worker training, work difficulty and so on. On the basis of these, we constructed their 3D CAD models and then drawn and reviewed their dismantling processes. In this study, the 3D simulation results of the shielding concrete and the RSR among main components are also presented and discussed.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.446-452
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• 2018

To improve the accuracy and safety of irradiation tests in High flux Advanced Neutron Application ReactOr (HANARO), the nuclear energy deposition rate, which is called nuclear heating, was estimated for an irradiation capsule with an iridium sample in the irradiation hole in order. The gamma rays emitted from the radioisotopes (RIs) of the structural materials such as flow tubes of fuel assemblies and heavy water reflector tank were considered as radiation source. Using the ORIGEN2.1 code, emission rates of delayed gamma rays were calculated in consideration of the activation procedure for 8 years and 2 months of HANARO operation. Calculated emission rates were used as a source term of delayed gamma rays in the MCNP6 code. By using the MCNP code, the nuclear heating rates of the irradiation capsules in the inner core, outer core, and heavy water reflector tank were estimated. Calculated nuclear heating in the inner core, outer core, and heavy water reflector tank were 200-260 mW, 80-100 mW, and 10 mW, respectively.

• Proceedings of the IEEK Conference
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• 2002.06a
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• pp.235-238
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• 2002

In this paper we describe the design and implementation of Electrical Ground Support Equipment(EGSE) for the CBS transponder testing. The main task of EGSE is to check out satellite systems, at system or subsystem level, during integration and validation phases of their life-cycle. Through a combination of hardware and software elements, EGSE supports manual, semi-automatic and fully automated testing. Automation is achieved by offering users simple, yet powerful means to write their own test application programs (test sequences) in high-level, test-oriented language and to run them in a strict real-time environment. The core of this environment is a user-configurable real-time database, containing all the information needed to calibrate acquired data, check them against predefined thresholds, automatically react to out-of-range conditions, display data using animated graphics or synoptic windows, and so on.

• Transactions of the KSME C: Technology and Education
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• v.5 no.2
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• pp.135-143
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• 2017

The domestic unique research reactor, HANARO (Hi-flux Advanced Neutron Application ReactOr), has been constructed with the open-pool, the core is submerged in, for the multi-purpose neutron application. The reactor has a primary cooling system to remove the fission heat from the core and its connected fluidic systems. Since the works are required at the reactor pool top as a characteristic of the research reactor, the radiation shall be minimized with the operation of the hot water layer system to avoid unnecessary radiation exposure on the workers during work at the pool top. Moreover, the pool water management system is connected to the reactor pool to maintain the pool temperature below $50^{\circ}C$ to minimize the uprising radioactive gas or impurity from the colder pool bottom. For the efficient flow rate of the PWMS, the thermal capacity of heat exchanger is selected with 260 kW in the normal operation condition. In this paper, the modeling is formulated to figure out whether or not each pool temperature maintains below the temperature limit and the calculation results show that the designed PWMS heat exchanger has enough capacity with the design margin regardless of the reactor operation mode.

• 한국신재생에너지학회:학술대회논문집
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• 2007.11a
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• pp.318-321
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• 2007

CdSe nanoparticles were prepared by chemical solution methods using $CdCl_2{\cdot}4H_2O$ (or $Cd(NO_3)_ 2{\cdot}4H_2O$) and $Na_2SeSO_3$. The characteristics of CdSe nanoparticles were controlled by the react ion time, reaction temperature and reaction method as well as the surfactants. Cetyltrimethyl ammonium bromide(CTAB) was used as a capping agent to control the chemical reactions in aqueous solution. Polyvinylalcohol(PVA) was used as a templet in sono-chemical method. CdSe nanoparticles synthesized in aqueous solution showed homogeneous size distribution with relatively stable surface. CdSe nanoparticles synthesized in non-aqueous solution containing diethanolamine(DEA) showed the structure transformation from cubic to hexagonal as the reduction temperature increased from 80 to $160^{\circ}C$. Core shell CdSe was synthesized by sono-chemical method. Characteristics of CdSe nanoparticles were analyzed using transmission electron microscopy(TEM), x-ray photoelectron spectroscopy(XPS), x-ray diffraction(XRD), UV-Vis absorption spectra, fourier transform infrared spectroscopy(FT-IR) and photoluminescence spectra spectroscopy(PL). This paper presents simple routes to prepare CdSe nanoparticles for solar cell applications.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.16-25
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• 2016

In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconiumalloy fuel claddingmaterials are rapidlyheateddue to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF) design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI) is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in $1,200-1,500^{\circ}C$ steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstratedcorrosionresistance.Asthese composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Moalloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are discussed in this document. In addition to assisting plants in meeting Light Water Reactor (LWR) challenges, accident-tolerant Mo-based cladding technologies are expected to be applicable for use in high-temperature helium and molten salt reactor designs, as well as nonnuclear high temperature applications.

• Journal of the Korea Society of Computer and Information
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• v.23 no.10
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• pp.103-109
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• 2018

In this paper, we propose a dynamic document classification method which breaks away from existing document classification method with artificial categorization rules focusing on suppliers and has changing categorization rules according to users' needs or social trends. The core of this dynamic document classification method lies in the fact that it creates classification criteria real-time by using topic modeling techniques without standardized category rules, which does not force users to use unnecessary frames. In addition, it can also search the details through the relevance analysis by calculating the relationship between the words that is difficult to grasp by word frequency alone. Rather than for logical and systematic documents, this method proposed can be used more effectively for situation analysis and retrieving information of unstructured data which do not fit the category of existing classification such as VOC (Voice Of Customer), SNS and customer reviews of Internet shopping malls and it can react to users' needs flexibly. In addition, it has no process of selecting the classification rules by the suppliers and in case there is a misclassification, it requires no manual work, which reduces unnecessary workload.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.952-964
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• 2016

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium-water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium-water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.