• Nuclear Engineering and Technology
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• v.16 no.1
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• pp.36-46
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• 1984

Studied are the statistical thermal design (STD) methods that have been developed to satisfy the design basis which protects a pressurized water reactor (PWR) core against departure from nucleate boiling (DNB) during normal operations and anticipated transients. The objective of the statistical thermal design is to quantify the thermal design margin and to remove any excess conservatism from the DNB ratio calculations through statistically combining design parameter uncertainties, while still maintaining a high level of core protection. This report describes and compares the STD methods developed by the two U.S. reactor vendors (Westinghouse and B & W). Included are the characteristics of STD, statistical treatment of uncertainties, DNB design limit development methodology and the sample application of the STD technique to core thermal design analysis. It is observed that the STD methods developed by the two vendors are similiar to each other in principle, but different in the treatment of the uncertainties associated with the design parameters. The statistical thermal design is found to significantly improve the thermal design margin.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.27 no.12
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• pp.652-658
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• 2015

The Bubble Jet Loop Heat Pipe (BJLHP) is designed to operate in the horizontal orientation. The motion of the bubble generated by boiling working fluid on a heater surface in the evaporating section of the BJLHP helps the working fluid transfer heat to the condensing portion. In this study, we changed the position of the heater in the evaporating section from concentric to eccentric. The concentric heater is located at the center of the tube in the evaporating part, and the eccentric heater is located at the bottom of the inner surface of the same tube. We used R-134a as the working fluid, and the charging ratio was 50%vol. We measured the temperatures of the evaporating and condensing sections by changing the input electric power from 50 W to 200 W, measuring every 50 W. The results of the experiment show that the effective thermal conductivity of BJLHP using an eccentric heater is four times higher than the BJLHP obtained using a concentric heater. Additionally, we conducted a visualization experiment on the evaporating portion of BJLHP to determine why the effective thermal conductivity was higher. The working fluid was water, and we took pictures of the flow visualization for BJLHP. Nucleate boiling with the eccentric heater was more intense and generated more bubbles. Therefore, the eccentric heater was more saturated by the liquefied working fluid.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2008.06a
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• pp.81-81
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• 2008

Applications of CdS films in this study are to exhibit a high conductivity when they are exposed at light with visible wavelength and sequentially to show a low conductivity in dark state. For this purpose, CdS films should have a high photosensitivity, still maintaining a high conductivity at a visible light. In this study, CdS films were prepared at room temperature on glass substrates by rf magnetron sputtering. In order to increase the photo-conductivity in visible light, various defect levels should be located within the CdS band gap. In order to nucleate the defect sites within the CdS band gap, CdS films were deposited on glass substrates at room temperature using various $H_2$/(Ar+$H_2$) flow ratios by an rf magnetron sputtering. Through the investigation of the structural and photoconductive properties of CdS films by an addition of hydrogen, the relationship between photo- and dark-resistance in CdS films was investigated in detail. 200-nm-thick CdS films for photoconductive sensor applications were prepared at various $H_2$/(Ar+$H_2$) flow ratios on glass substrates at room temperature by rf magnetron sputtering. Sulfur concentration in CdS films crystallized at room temperature with (002) preferred orientation depends directly on the hydrogen atmosphere and the surface roughness of the films gradually increases with increasing hydrogen atmosphere. Films deposited at 8% of $H_2$/(Ar+$H_2$) exhibit an abrupt decrease of dark- and photo-resistance, showing a low photo-sensitivity ($R_{dark}/R_{photo}$). Onthe other hand, films deposited at a hydrogen atmosphere of 42% exhibit a photo-sensitivity of $5\times10^3$, maintaining a photo-resistance of an approximately $2\times10^4\Omega$/square. The dark- and photo-resistance values of CdS films were related with a composition, surface roughness, and defect sites within the band gap.

• Journal of Radiation Protection and Research
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• v.17 no.1
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• pp.1-10
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• 1992

A number of investigators have reported the formation of the radiolytic ultrafine particles produced by the interaction of ionizing radiation with water vapor. Previous studies have suggested that a very high localized concentration of the OH radical produced by the radiolysis of water can react with trace gas like organic vapors and produce lower vapor pressure compounds that can then nucleate. In order to determine water vapor dependence of the active, positively charged, first radon daughter(Po-218), an experiment was conducted using a well-controlled radon chamber. The activity size distribution of the radon daughter in the range of 0.5-100nm was measured using the parallel graded wire screens system. Measurements were taken for different relative humidity. The resultant activity size distributions were analyzed. The addition of water vapor to the radon carrier gases resulted in the formation of ultrafine particles by OH radicals formed by radon radiolysis. It may be due to the neutralization of charged Po-218 ion with water vapor through the radio lysis.

• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.528-539
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• 2004

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.

• Nuclear Engineering and Technology
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• v.14 no.3
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• pp.122-137
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• 1982

Analyzed is the sensitivity of reactor transient behavior to various reactor parameters during the reactivity induced accidents (RIA) of the Kori Unit 1. Included in the analysis is a partial spectrum of RIAs with relatively fast transients such as uncontrolled rod cluster control assembly bank withdrawl from a subcritical or low power startup condition and rod ejection accidents. The analysis can be performed generally in three steps: calculation of an average core power change, hot spot heat transfer calculation and DNBR (departure from nucleate boiling ratio) calculation. The computer codes used for the analysis are either developed based on the codes relevent to it. These codes are evaluated to be highly reliable. An extensive sensitivity analysis is performed to study the effects of various reactor design and operating parameters on the reactor transient behavior during the accidents. The assumptions and initial conditions used for the RIA analysis in the Kori Unit 1 FSAR (Final Safety Analysis Report) are reexamined, and the corresponding analysis results are reassessed, based on the sensitivity analysis results, to be conservative and reliable.

• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.236-247
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• 1983

Analyzed are the the DNBR(Departure from Nucleate Boiling Ratio) sensitivities to variations in various PWR operating parameters utilizing the Korea Nuclear Unit 1(KNU-1) design and operating data. Studied parameters in the analysis are core power level, system pressure, core inlet flow rate, core inlet temperature, enthalpy rise hot channel factor, and axial power peaking factor and axial offset. The calculations are performed using the steady state and transient thermal-hydraulics computer program, COBRA-IV-K, which is the revised version of COBRA-IV-i that has been adapted, partially modified and verified at KAERI. A reference case is established based on the design and operating condition of the KNU-1 reactor core, and this provides a basis for the subsequent sensitivity analysis. From the calculation results it is concluded that the most sensitive parameter in the DNBR thermal design is the coolant core inlet temperature while the axial power peaking factor is the least sensitive.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.19 no.9
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• pp.615-621
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• 2007

Experimental study on the heat transfer characteristics of $CO_2$ in a horizontal smooth tube was carried out to investigate the heat transfer coefficient and pressure drop during evaporation of $CO_2$. The experiment apparatus consisted of a test section, a DC power supply, a heater, a chiller, a mass flow meter, a pump and a measurement system. Experiment was conducted for various mass fluxes ($200{\sim}1200kg/m^2s$), heat flukes ($10{\sim}100kW/m^2$) and saturation temperatures (-5, 0, $5^{\circ}C$). With increasing the heat flux, the evaporation heat transfer coefficient increased. But the variation of the heat transfer coefficient on the increase of the mass flux was not large. And the significantly drops of the heat transfer coefficient was observed at any heat flux and mass flux because of the change of the flow pattern in the tube. With increasing the saturation temperature, the heat transfer coefficient increased due to the promotion of a nucleate boiling. The measured pressure drop during evaporation increased with increasing the mass flux and decreasing the saturation temperature.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.22 no.7
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• pp.980-991
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• 1998

In this study, nucleate pool boiling heat transfer coefficients of alternative refrigerants on a plain, low fin, and two enhanced tubes were measured and compared against each other. To obtain data at conditions similar to the actual evaporator, a fluid heating method was employed instead of an electric heating method in the experiments. R123, R134a, R22 and R410a were used as working fluids and data were taken at 7 deg.C ar heat fluxes of 20 ~ 100 kW/m$\^$2/. Comparison of the plain tube data against some correlations showed that the simplest correlation of Cooper based on reduced pressure predicted the data for all fluids tested with a 10% deviation. For all refrigerants, enhanced tubes composed of subsurface and subtunnels, especially Thermoexcel-E tube, showed the highest heat transfer coefficients among the tubes tested with one exception that the low fin tube's performance was better than those of enhanced tubes for high vapor pressure fluid such as R410a at high heat flux. Finally, a low fin and enhanced tubes showed higher heat transfer enhancement for low vapor pressure of R123 than for high vapor pressure fluisd. For R123, the enhancement factors for Turbo-B and Thermoexcel-E tubes were 2.8 ~ 4.8 and 4.6 ~ 8.1 respectively.

• Journal of Advanced Marine Engineering and Technology
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• v.29 no.3
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• pp.297-305
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• 2005

The evaporation heat transfer coefficient of $CO_2$ (R-744) in a horizontal tube was investigated experimentally. The experiments were conducted without oil in a closed refrigerant loop which was driven by a magnetic gear pump. The main components of the refrigerant loop are a receiver. a variable-speed pump. a mass flow meter. a pre-heater and evaporator (test section). The test section consists of a smooth. horizontal stainless steel tube of 7.75 mm inner diameter. The experiments were conducted at mass flux of 200 to $500\;kg/m^{2}s$. saturation temperature of $-5^{\circ}C\;to\;5^{\circ}C$. and heat flux of 10 to $40\;kW/m^2$. The test results showed the evaporation heat transfer of $CO_2$ has greatly effect on more nucleate boiling than convective boiling. The evaporation heat transfer coefficients of $CO_2$ are highly dependent on the vapor quality. heat flux and saturation temperature. The evaporation heat transfer coefficient of $CO_2$ is very larger than that of R-22 and R-134a. In making a comparison between test results and existing correlations. the present experimental data are the best fit for the correlation of Jung et al. But it was failed to predict the evaporation heat transfer coefficient of $CO_2$ using by the existing correlation. Therefore. it is necessary to develop reliable and accurate predictions determining the evaporation heat transfer coefficient of $CO_2$ in a horizontal tube.