• Korean Journal of Food Science and Technology
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• v.4 no.1
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• pp.6-12
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• 1972

Fay Elberta freestone peaches were harvested at four maturity levels as judged by skin color and firmness. They were ripened at $68^{\circ}F$ under 80 percent relative humidity for 4, 6, 8 and 10 days respectively prior to canning. Results indicate that both harvest maturity and ripening conditions are important factors influencing quality, flavor and composition of canned freestone peaches. Peaches harvested at $18{\sim}24$ pounds on a Magness-Taylor pressure tester with a 7/16' plunger(M1) failed to ripen satisfactorily. Fruits harvested at $13{\sim}17$ pounds (M2) pressure test ripened successfully at $68^{\circ}F$ within 6 to 8 days; and those harvested at 6 to 12 pounds (M3) needed 4 days for ripening at $68^{\circ}F$. Tree-ripened fruits (M4) were undesirable for canning because of the high percentage of bruised fruits. The optimum firmness for canning appears to be in the range of 1.5 to 3.0 pounds. The titratable acidity of peaches decreased during maturation and ripening. The tannin content of peaches at M1 maturity decreased with ripening at $68^{\circ}F$. But no appreciable change was observed in the M2 and M3 series which were ripened at $68^{\circ}F$ for 4 to 10 days. The volatile reducing substances (V.R.S.) increased as the peaches developed on the tree and also during post-harvest ripening. The effect of harvest maturity and post-harvest ripening on color grade of the canned peaches is presented. Little difference was found in the flavor and composition of peaches peeled by the cup-down lye peeling and the steam-peeling methods. The cup-down lye-peeling method might be most advantageous because of its higher peeling efficiency.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.20-27
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• 2010

In order to prevent the corrosion of component contacted primary water designed alloy 600 material in the nuclear power plant. But the primary water stress corrosion cracking(PWSCC) of alloy 600 and weld area occurs continuously due to the residual stress. The leakage accident resulted from PWSCC in the drain nozzle of the steam generator of domestic power plants. Heater sleeves of the pressurizer are welded with alloy 600 weld material and therefore exposed to the primary water environment. PWSCC occurred in heater sleeve material and weld area of many foreign power plants. The current issue of domestic nuclear power plants are consequently concentrated to PWSCC of similar material. In order to improve the detection and the sizing of the PWSCC in the welding sleeve of the pressurizer, the automatic UT system and multi-directions probe sets have been developed. The experimental studies have been performed using the mock-up block containing artificial reflectors(ID connected EDM notch) and semi-artificial cracks made from thermal fatigue. The automatic UT System is applied in the detection and the length sizing of the ID/OD on the tube and the J-groove weld area of the artificial reflectors and results of the detection and the sizing are compared respectively. Also, the developed automatic UT system is successfully accomplished to inspect the heater sleeve and the J-groove weld area on the pressurizer for the detection of PWSCC.

• Journal of ILASS-Korea
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• v.5 no.3
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• pp.17-26
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• 2000

The main parameter commonly used to evaluate spray distribution is spray angle. Spray angle is important because it influences the axial and radial distribution of the fuel. Spray angles were measured and compared for the two non-air assisted injectors such as 2hole-2stream 4hole-1stream injectors used for port fuel injection gasoline engines with n-heptane as a fuel by three different measuring techniques, i.e., digital image processing, shadowgraphy, and spray patternator, respectively. Fuel was injected with the injection pressures of 0.2-0.35 MPa into the room temperature and atmospheric pressure environment. In digital image processing approach, the selection of the transmittance level is critical to obtain the edge of spray and hence to measure the spray angle. From the measurement results by the shadowgraphy technique, it is dear that the spray angle is varied during the spray injection period. The measurement results from spray patternator show that the different spray angles exist in different region. Spray angle increases with the increase in the injection pressure. it is suggested that the spray angle and stream separated angle should be specified when spray is characterized for 2hole-2stream injector, because spray angle is much different though stream separated angle is same. It was also considerably affected by the measurement techniques introduced in this experimental work. However, the optimal axial distance for measuring the spray angle seems to be at least 60-80 mm from the injector tip for two non-air assisted injectors.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.274-284
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• 1992

A 1/15-scale CREARE experiment, which simulates the thermal-hydraulic behavior in the reactor pressure vessel of a PWR during a hypothetical Loss Of Coolant Accident, has been analyzed using CATHARE code for the associated model assessment to represent the phenomenon. The key parameters examined in the CREARE experiment were known as ECC water injection rate. ECC water subcooling, system pressure, and steam flow rate coming out from the core bottom. The present CATHARE simulation, however, has been mainly focused on qualitative analysis of a countercurrent flow in the downcomer. The discrepancy of the simulation results with the experimental data is considered arising primarily from an inadequate numerical representation as well as an interfacial friction model. Accordingly it is suggested from the sensitivity studies that either multidimensional approach or further examination of momentum equations at a junction near a volume element in CATHARE be necessary in order to represent the phenomenon more realistically.

• Proceedings of the Korea Concrete Institute Conference
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• 2008.04a
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• pp.441-444
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• 2008

Ultra high performance concrete (UHPC) in this study is composed of sand, cement, silica fume, siliceous powder, superplasticizer and steel fiber. UHPC is composed of fine mineral particles below 0.5mm in diameter. In general, siliceous powder improves the mechanical properties of concrete by physical and chemical effect. Physical effect is related with filling interior voids which weaken the mechanical properties and chemical effect with reaction of $SiO_2$ with cement hydrates in a condition of high temperature and pressure. We evaluated the effect of siliceous powder's particle size on the mechanical properties of ultra high performance concrete in air pressure and $90^{\circ}C$ steam curing condition. siliceous powder's particle size in this study is in the range of $2{\mu}m$ to $26{\mu}m$. Fluidity in a fresh concrete, compressive strength, ultimate strain, elastic modulus and flexural strength in a hardened concrete was evaluated. We could find out that the smaller siliceous powder's particle size is, the better the fluidity and strength properties.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.12-19
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• 1990

The Loss of Offsite Power Transient at 77.5% power which occurred on June 9, 1981 at the Kori Unit 1 PWR (Pressurized Water Reactor) is simulated using the RELAP5/MOD2 system thermal-hydraulics computer code. Major thermal-hydraulic parameters are compared with the available plant data. The comparison of the analysis results with the plant data demonstrates that the RELAP5/MOD2 code has the capability to simulate the thermal-hydraulic behaviour of PWRs under accident conditions of this type with accuracy, except the pressurizer pressure and level. The pressurizer pressure increase is sensitive to the in surge now it is believed that the interracial heat transfer in a horizontal stratified flow regime may be estimated low and the compression effect due to insurge flow may be high. In the nodalization sensitivity study it is found that S/G noding with junctions between bypass plenum and steam dome is preferred to simulate the S/G water level decreasing and avoid the spurious level peak at trubine trip.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1066-1080
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• 2024

SMART100 has a containment pressure and radioactivity suppression system (CPRSS) for passive containment cooling system (PCCS). This prevents overheating and over-pressurization of a containment through direct contact condensation in an in-containment refueling water storage tank (IRWST) and wall condensation in a CPRSS heat exchanger (CHX) in an emergency cool-down tank (ECT). The Korea Atomic Energy Research Institute (KAERI) constructed scaled-down test facilities, SISTA1 and SISTA2, for the thermal-hydraulic validation of the SMART100 CPRSS. Three separate effect tests were performed using SISTA1 to confirm the heat removal characteristics of SMART100 CPRSS. When the low mass flux steam with or without non-condensable gas is released into an IRWST, the conditions for mitigation of the chugging phenomenon were identified, and the physical variables were quantified by the 3D reconstruction method. The local behavior of the non-condensable gas was measured after condensation inside heat exchanger using a traverse system. Stratification of non-condensable gas occurred in large tank of the natural circulation loop. SISTA2 was used to simulate a small break loss-of-coolant accident (SBLCOA) transient. Since the test apparatus was a metal tank, compensations of initial heat transfer to the material and effect of heat loss during long-term operation were important for simulating cooling performance of SMART100 CPRSS. The pressure of SMART100 CPRSS was maintained below the design limit for 3 days even under sufficiently conservative conditions of an SBLOCA transient.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2232-2237
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• 2007

Steam reforming and catalytic reforming of $CH_4$ conversion to produce synthesis gas require both high temperatures and high pressure. Non-thermal plasma is considered to be a promising technology for the hydrogen rich gas production from methane. In this study, three phase AC GlidArc plasma system was employed to investigate the effects of gas composition, gas flow rate, catalyst reactor temperature and applied electric power on the $CH_4$ and $H_2$ yield and the product distribution. The studied system consisted of three electrode and it connected AC generate power system different voltages. In this study, air was used for the partial oxidation of methane. The results showed that increasing gas flow rate, catalyst reactor temperature, or electric power enhanced $CH_4$ conversion and $H_2$ concentration. The reference conditions were found at a $O_2$/C molar ratio of 0.45, a feed flow rate of 4.9 ${\ell}$/min, and input power of 1kW for the maximum conversions of $CH_4$ with a high selectivity of $H_2$ and a low reactor energy density.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.3458-3463
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• 2007

Leakage would happen because of the damage of high temperature and high-pressure valve in nuclear power plant. condition based prevention maintenance is essential by using the suitable method based on local condition. Energy loss prevention can prevent from an accurate test, Local actually and ability. The methods of test for high energy fluid leakage at present are analysis of ${\Delta}$T, AE(Acoustic Emission) analysis, and thermal image. The result for test of AC (Main steam) system in YNG 2 Unit reveals that the AE occurred clearly in leakage situation, but thermal image didn't occur. It is identified that leakage is occurred when the orifice located front and back of valve operates. It shows that making a impatient judgment by using the single method if it is leakage is containing uncertainty. So I think that using the Multi-Measuring method is more sound judgment than Single-Measuring method.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.8
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• pp.1645-1652
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• 2002

The failure analysis on fractured parts is divided into the qualitative method by naked eyes and metallurgical microscope etc. and the quantitative method by SEM and X-ray diffraction etc. X-ray fractography can be applied to contaminated surface as well as clean surface and gain the plastic deformation and the residual stress near the fractured surface. Turbine blade is subject to cyclic bending force by steam pressure and suffers fatigue damage according to the increasing operating time. Therefore, to clean up the fracture mechanism of torsion-mounted blade in nuclear plant, the fatigue and the X-ray diffraction test was performed on the 12%Cr steel fur turbine blade and the fractured parts. The correlation of X-ray parameter and fracture mechanics parameter was determined, and then the load applied to actual broken turbine blade was predicted. Failure analysis was performed by contact stress analysis and Goodman diagram of torsion-mounted blade.