• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.357-364
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• 2022

The safety of a KTC-360 transport cask, a large-capacity pressurized heavy-water reactor transport cask that transports CANDU spent nuclear fuel discharged from the reactor after burning in a pressurized heavy-water reactor, must be demonstrated under the normal transport and accident conditions specified under transport cask regulations. To confirm the thermal integrity of this cask under normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62℃, indicating that such casks must be transported separately. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC-360 cask can be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel were 446℃ lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fire accident conditions.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2340-2345
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• 2008

Cylindrical stainless-steel/sodium heat pipe for a high-temperature application was manufactured and tested for transient and steady-state operations. Two layers of Stainless-steel screen mesh wick was inserted as a capillary structure. The outer diameter of the heat pipe was 12.7 mm and the total length was 250 mm. As thermal transport conditions, the effective transport length, the heat flux, the tilt angle and the operating temperature were varied. The heat was supplied by an electric furnace up to 1 kW and the cooling was performed by forced convection of air. The effective thermal conductivity and the thermal resistance were investigated as a function of heat flux, heat transport length, and vapor temperature. Typical range of the total thermal resistance was as low as $0.036^{\circ}C/W$ at $175.8\;kW/m^2$ of heat flux and $700^{\circ}C$ of operating temperature.

• 한국태양에너지학회:학술대회논문집
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• 2008.11a
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• pp.220-225
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• 2008

Cylindrical stainless-steel/sodium heat pipe for a high-temperature solar thermal application was manufactured and tested for transient and steady-state operations. Two layers of stainless-steel screen mesh wick was inserted as a capillary structure. The outer diameter of the heat pipe was 12.7 mm and the total length was 250 mm. The effective heat transport length, the thermal load, and the operating temperature were varied as thermal transport conditions of the heat pipe. The thermal load was supplied by an electric furnace up to 1kW and the cooling was performed by forced convection of air The effective thermal conductivity and the thermal resistance were investigated as a function of heat flux, heat transport length, and vapor temperature. Typical range of the total effective thermal conductivity was as low as 43,500 W/m K for heat flux of 176.4 kW/$m^2$ and of operating temperature of 1000 K.

• Nuclear Engineering and Technology
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• v.29 no.4
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• pp.269-279
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• 1997

The errors associated with incorrect sampling and transport of radioactive aerosol from radwaste thermal process off-gas are analyzed and the conditions of representative sampling and correct transport of radioactive aerosol for off-gas system evaluation are discussed. An estimation method of sampling errors for individual radionuclides is proposed and applied to simulated vitrification melter aerosols. Prediction methods for particle deposition in sample transport tube under laminar as well as turbulent flow conditions are also described by example calculations with simulated incinerator off-gas From the results of example calculations and plots, instrumental and operational conditions of radioactive aerosol sampling system with minimized errors and correction methods for nonideal sampling and transport are recommended.

• Journal of ILASS-Korea
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• v.27 no.2
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• pp.66-76
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• 2022

This study presents a prediction methodology of transport properties using the methane-based TRAPP (m-TRAPP) method in a wide range of temperature and pressure conditions including both subcritical and supercritical regions, in order to obtain thermo-physical properties for hydrocarbon aviation fuels and their products resulting from endothermic reactions. The viscosity and thermal conductivity are predicted in the temperature range from 300 to 1000 K and the pressure from 0.1 to 5.0 MPa, which includes all of the liquid, gas, and the supercitical regions of representative hydrocarbon fuels. The predicted values are compared with those data obtained from the NIST database. It was demonstrated that the m-TRAPP method can give reasonable predictions of both viscosity and thermal conductivity in the wide range of temperature and pressure conditions studied in this paper. However, there still exists large discrepancy between the current data and established values by NIST, especially for the liquid phase. Compared to the thermal conductivity predictions, the calculated viscosities are in better agreement with the NIST database. In order to consider a wide range of conditions, it is suggested to select an appropriate method through further comparison with another improved prediction methodologies of transport properties.

• Journal of Environmental Science International
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• v.10 no.2
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• pp.143-151
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• 2001

The atmospheric conditions and the transport mechanism of long-range transport of air pollutants from coastal area to inland area were investigated using regular meteorological data and air pollution data obtatined from the southeastern area of Korea. Daytime temperature over the inland area(Taegu) was higher than that over the coastal area(Pusan) and the temperature difference of about 5~6$^{\circ}C$ when the thermal low most fully developed and the sea level pressure over Taegu was lower than that over Pusan by about 4~5hPa at that time. Therefore this low pressure appeared to the thermally induced low. Air mass polluted from the coastal area during the morning period was transported inland area, at first by the sea breeze and by the large scale wind system toward the thermal low generated in the mountainous inland region. This was explained by the fact that the concentration of air pollutants over Taegu increased throughtout the late afternoon.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.161-167
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• 2011

In case that the maximum temperature of any surface readily accessible during transport of a spent nuclear fuel (SNF) transport cask exceeds $85^{\circ}C$ in the absence of insolation under the ambient temperature of $38^{\circ}C$, personnel barriers or transport hood shall be used to prevent people from casual contact with the transport cask surface. Usually the air temperature within the hood and the hood surface temperature are calculated and further utilized as boundary conditions(free stream temperature and external radiation temperature) for thermal evaluation under normal conditions of transport. In this study, these temperatures are derived using the analytical method based on the heat transfer mechanism around the transport cask under transport hood assuming the thermal equilibrium. By comparing the analytical solutions with the results from the detailed calculations with CFD-computer-code FLUENT 12.1 it is verified that the analytical method is still efficient tool to estimate the temperatures and these temperatures can be further used as boundary conditions for thermal evaluation under normal conditions of transport.

• Journal of Radiation Protection and Research
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• v.28 no.4
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• pp.281-290
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• 2003

The KN-12 spent nuclear fuel transport cask, which is a Type B(U) package designed to comply with the requirements of Korea Atomic Energy Act[1], IAEA Safety Standards Series No.TS-R-1[2] and US 10 CFR Part 71[3], is designed for carrying up to 12 PWR spent fuel assemblies in a basket structure. The cask has been licensed in accordance with Korea Atomic Energy Act and was fabricated in Korea in accordance with the requirements of ASME B&PV Sec.III, Div.3[4]. The cask must maintain thermal integrity in accordance with the related regulations and be evaluated to verify that the thermal performance of the cask complies with the regulatory requirements. The temperatures of the cask and components were determined by using finite elements methods with a numerical tool, safety tests using an 1/8 height slice model of the real cask were conducted to demonstrate verification of the numerical tool and methods, and heat transfer tests for normal transport conditions were performed as a fabrication acceptance test to demonstrate the heat transfer capability of the cask.

• Journal of the Korean Crystal Growth and Crystal Technology
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• v.32 no.1
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• pp.16-24
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• 2022

A computational study of combined thermal and solutal convection (double diffusive convection) in a sealed crystal growth reactor is presented, based on a two-dimensional numerical analysis of the nonlinear and strongly coupled partial differential equations and their associated boundary conditions. The average Nusselt numbers for the source regions are greater than those at the crystal regions for 9.73 × 10³ ≤ Gr_t ≤ 6.22 × 10⁵. The average Nusselt numbers for the source regions varies linearly and increases directly with the thermal Grashof number form 9.73 × 10³ ≤ Gr_t ≤ 6.22 × 10⁵ for aspect ratio, Ar (transport length-to-width) = 1 and 2. Additionally, the average Nusselt numbers for the crystal regions at Ar = 1 are much greater than those at Ar = 2. Also, the occurrence of one unicellular flow structure is caused by both the thermal and solutal convection, which is inherent during the physical vapor transport of Hg₂Br₂. When the aspect ratio of the enclosure increases, the fluid movement is hindered and results in the decrease of thermal buoyancy force.

• Journal of Radiation Protection and Research
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• v.22 no.4
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• pp.251-261
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• 1997

A package to transport the high level radioactive materials is required to withstand the hypothetical accident conditions as well as normal transport conditions according to IAEA standards and domestic regulations. The regulations require that the package should maintain the shielding, thermal and structural integrities to release no radioactive material. In general, safety evaluation of packages is performed by experimental methods using scale model and/or analytical methods using computer codes. This paper presents the safety evaluation of package to transport the radioisotopes produced in the HANARO to the radioisotope production facility. Radiation shielding, thermal and structural analyses were peformed using the computer codes. It has been verified that the package is safe under hypothetical accident conditions as well as normal transport conditions.