• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.2
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• pp.125-132
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• 2012

The dehydration schemes of rare earth (La, Ce, Nd, Pr, Sm. Eu, Gd, Y) chloride hydrates was investigated by using a dehydration apparatus. To prevent the formation of the rare earth oxychlorides, the operation temperature was changed step by step ($80{\rightarrow}150{\rightarrow}230^{\circ}C$) based on the TGA (thermo-gravimetric analysis) results of the rare earth chloride hydrates. A vacuum pump and preheated Ar gas were used to effectively remove the evaporated moisture and maintain an inert condition in the dehydration apparatus. The dehydration temperature of the rare earth chloride hydrate was increased when the atomic number of the rare earth nuclide was increased. The content of the moisture in the rare earth chloride hydrate was decreased below 10% in the dehydration apparatus.

• Journal of the Korean Solar Energy Society
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• v.38 no.2
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• pp.55-66
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• 2018

South Korea aims to shift the 20 percent of electricity supplement from the fossil fuel including the nuclear to renewable energy systems by 2030. In order to realize this agenda in the buildings, the plus energy house is necessary to increase the renewable energy supplement beyond the zero energy house. This paper suggested KePSH (KIER Energy-Plus Solar House) and energy performance of house and renewable energy systems was investigated. The KePSH has the target of generating 40% surplus energy than the conventional house energy consumption. The plus energy house is the house that generates surplus energy from the renewable energy sources than that consumes. In order to minimize the cooling and heating load of the house, the shape design and passive parameters design were conducted. Based on the experimental data of the plug load in the typical house, the total energy consumption of the house was estimated. This paper also suggested renewable energy sources integrated HVAC system using air-source heat pump system. Two cases of renewable energy system integration methods were suggested, and energy performance of the cases was investigated using TRNSYS 17 program. The results showed that the BIPV (building integrated photovoltaic) system (i.e., CASE 1) and BIPV and BIST system (i.e., CASE 2) shows 42% and 29% of plus energy rate, respectivey. Also, CASE 1 can generate 59% more surplus energy compared with the CASE 2 under the same installation area.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.32 no.2A
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• pp.123-130
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• 2012

A cooling system is indispensable for the fossil and nuclear power plants which produce electricity by rotating the turbines with hot steam. A cycle of the typical cooling system includes pumping of seawater at the intake pump house, exchange of heat at the condenser, and discharge of hot water to the sea. The cooling type of the nuclear power plants in Korea recently evolves from the conventional surface intake/discharge systems to the submerged intake/discharge systems that minimize effectively an intake temperature rise of the existing plants and that are beneficial to the marine environment by reducing the high temperature region with an intensive dilution due to a high velocity jet and density differential at the mixing zone. It is highly anticipated that the future nuclear power plants in Korea will accommodate the submerged cooling system in credit of supplying the lower temperature water in the summer season. This study investigates the approach flow patterns at the velocity caps and discharge flow patterns from diffusers using the 3-D computational fluid dynamics code of $FLOW-3D^{(R)}$. The approach flow test has been conducted at the velocity caps with and without a cap. The discharge flow from the diffuser was simulated for the single-port diffuser and multi-ports diffuser. The flow characteristics to the velocity cap with a cap demonstrate that fish entrainment can significantly be minimized on account of the low vertical flow component around the cap. The flow pattern around the diffuser is well agreed with the schematic diagram by Jirka and Harleman.

• Journal of Food Hygiene and Safety
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• v.31 no.4
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• pp.286-293
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• 2016

This study is an experiment for gastric protective effects of ursolic acid. In order to identify the effects of ursolic acid on gastrointestinal disorder, acute and chronic gastritis were also observed using HCl ethanol and indomethacin-induced gastric lesion models, respectively. As for gastric acid, it was also identified through proton pump ($H^+/K^+-ATPase$) inhibiting activity. In regards to protective factor for gastric damage, prostaglandin $E_2$ ($PGE_2$) was quantitatively analyzed. Antibacterial activity experiment was done on Helicobacter pylori (H.pylori), which is known to be the causing factor of chronic gastritis, gastric ulcer and gastric cancer. By making use of AGS cell, it was confirmed that ursolic acid was involved in apoptosis of gastric cancer cell through 4',6-diamidino-2-phenylindol (DAPI) staining and flow cytometry analysis. As a result, ursolic acid reduced gastric lesions caused by HCl ethanol and indomethacin. Ursolic acid inhibited acid secretion by inhibiting proton pump ($H^+/K^+-ATPase$), which is the gastric acid secreting enzyme involved at the final phase of gastric acid secretion. And ursolic acid was identified with gastric mucosa protection effects by increasing the concentration of $PGE_2$, a protective factor of gastric mucosa preservation. The antibacterial activity on H. pylori, which is aggressive factor in gastrointestinal disorder, ursolic acid showed inhibitory effects on H. pylori colonization. In the DAPI nuclear staining, unlike the control group, shape of the nucleus has deformed, and has been observed either shrinked cell or chromatin condensation phenomenon. In the Flow cytometry assay, confirmed the growth rate of apoptosis in a concentration-dependent manner.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.363-372
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• 2014

In this study, we proposed a newly designed sulfuric acid transfer system for the sulfur-iodine (SI) thermochemical cycle. The proposed sulfuric acid transfer system was evaluated using a computational fluid dynamics (CFD) analysis for investigating thermodynamic/hydrodynamic characteristics and material properties. This analysis was conducted to obtain reliable continuous operation parameters; in particular, a thermal analysis was performed on the bellows box and bellows at amplitudes and various frequencies (0.1, 0.5, and 1.0 Hz). However, the high temperatures and strongly corrosive operating conditions of the current sulfuric acid system present challenges with respect to the structural materials of the transfer system. To resolve this issue, we designed a novel transfer system using polytetrafluoroethylene (PTFE, $Teflon^{(R)}$) as a bellows material for the transfer of sulfuric acid. We also carried out a CFD analysis of the design. The CFD results indicated that the maximum applicable temperature of PTFE is about 533 K ($260^{\circ}C$), even though its melting point is around 600 K. This result implies that the PTFE is a potential material for the sulfuric acid transfer system. The CFD simulations also confirmed that the sulfuric acid transfer system was designed properly for this particular investigation.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2008.04a
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• pp.425-430
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• 2008

To verify the system performance of portable AE leak diagnosis system which can measure with moving conditions, AE activities such as RMS voltage level, AE signal trend, leak rate degree according to AE database, FFT spectrum were measured during operation on total 11 valves of the secondary system in nuclear power plant. AE activities were recorded and analyzed from various operating conditions including different temperature, type of valve, pressure difference, valve size and fluid. The results of this field study are utilized to select the type of sensors, the frequency band for filtering and thereby to improve the signal-to-noise ratio for diagnosis for diagnosis or monitoring of valves in operation. As the final result of application study above, portable type leak diagnosis system by AE was developed. The outcome of the study can be definitely applied as a means of the diagnosis or monitoring system for energy saving and prevention of accident for power plant valve. The purpose of this study is to verify availability of the acoustic emission in-situ monitoring method to the internal leak and operating conditions of the major valves at nuclear power plants. In this study, acoustic emission tests are performed when the pressurized temperature water and steam flowed through glove valve(main steam dump valve) and check valve(main steam outlet pump check valve) on the normal size of 12 and 18 ". The valve internal leak monitoring system for practical field was designed. The acoustic emission method was applied to the valves at the site, and the background noise was measured for the abnormal plant condition. To improve the reliability, a judgment of leak on the system was used various factors which are AE parameters, trend analysis, frequency analysis, voltage analysis and amplitude analysis of acoustic signal emitted from the valve operating condition internal leak.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.735-744
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• 2012

In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP ($\underline{A}$PR+ $\underline{C}$ore Flow & $\underline{P}$ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.

• Nuclear Engineering and Technology
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• v.41 no.10
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• pp.1263-1274
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• 2009

Several experimental tests to simulate a reflood phase of a cold-leg LBLOCA of the APR1400 have been performed using the ATLAS facility. This paper describes the related experimental results with respect to the thermal-hydraulic behavior in the core and the system-core interactions during the reflood phase of the cold-leg LBLOCA conditions. The present descriptions will be focused on the LB-CL-09, LB-CL-11, LB-CL-14, and LB-CL-15 tests performed using the ATLAS. The LB-CL-09 is an integral effect test with conservative boundary condition; the LB-CL-11 and -14 are integral effect tests with realistic boundary conditions, and the LB-CL-15 is a separated effect test. The objectives of these tests are to investigate the thermal-hydraulic behavior during an entire reflood phase and to provide reliable experimental data for validating the LBLOCA analysis methodology for the APR1400. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results for the LBLOCA scenario of the APR1400. The ECC water flow rate from the safety injection tanks and the decay heat were simulated from the start of the reflood phase. The simulated core power was controlled to be 1.2 times that of the ANS-73 decay heat curve for LB-CL-09 and 1.02 times that of the ANS-79 decay curve for LB-CL-11, -14, and -15. The simulated ECC water flow rate from the high pressure safety injection pump was 0.32 kg/s. The present experimental data showed that the cladding temperature behavior is closely related to the collapsed water level in the core and the downcomer.

• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.633-640
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• 2014

In this research, the surface roughness affecting the pressure drop in a pipe used as the steam generator of a PWR was studied. Based on the CFD (Computational Fluid Dynamics) technique using a commercial code named ANSYS-FLUENT, a straight pipe was modeled to obtain the Darcy frictional coefficient, changed with a range of various surface roughness ratios as well as Reynolds numbers. The result is validated by the comparison with a Moody chart to set the appropriate size of grids at the wall for the correct consideration of surface roughness. The pressure drop in a full-scale U-shaped pipe is measured with the same code, correlated with the surface roughness ratio. In the next stage, we studied a reduced scale model of a U-shaped heat pipe with experiment and analysis of the investigation into fluid-structure interaction (FSI). The material of the pipe was cut from the real heat pipe of a material named Inconel 690 alloy, now used in steam generators. The accelerations at the fixed stations on the outer surface of the pipe model are measured in the series of time history, and Fourier transformed to the frequency domain. The natural frequency of three leading modes were traced from the FFT data, and compared with the result of a numerical analysis for unsteady, incompressible flow. The corresponding mode shapes and maximum displacement are obtained numerically from the FSI simulation with the coupling of the commercial codes, ANSYS-FLUENT and TRANSIENT_STRUCTURAL. The primary frequencies for the model system consist of three parts: structural vibration, BPF(blade pass frequency) of pump, and fluid-structure interaction.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1083-1095
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• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.