• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.23 no.7
• /
• pp.856-863
• /
• 1999

The thermal performance degradation of heat pipes is caused by the non-condensable gas generation mainly due to the electrochemical corrosion which results from the reaction of working fluids with tube materials. In this study, a simplified method described below was proposed to estimate the life of heat pipes concerning the non-condensable gas generation. The temperature distributions at the outer surface of heat pipes was measured, and based on them the amount of non-condensable gas of hydrogen was estimated. Applying it to the Arrhenius model, the mass generation of hydrogen and the volume occupied by the gas In heat pipes could be estimated for an operating temperature and time. Moreover, this simplified method was applied to the accelerated life test of nine methanol-stainless steel heat pipe samples.

• Journal of Powder Materials
• /
• v.9 no.1
• /
• pp.50-60
• /
• 2002

SPS(Spark Plasma Sintering ) is known to be an excellent sintering method for porous materials. In the present work an attempt has been made of fabricating porous 316L Stainless steel with good mechanical properties by using controlled SPS process Porosity was 21%~53% at sintering temperature of $600^{\circ}C$~100$0^{\circ}C$ The limit of porosity with available mechanical strength was 30% at given experimental conditions. Porosity can be controlled by manipulating the intial height of the compact by means of the supporter and punch length. The applied pressure can be exerted entirely upon the supporter, giving no influence on the specimen. The specimen is then able to be sintered pressurelessly. In this case porosity could be controlled from 38 to 45% with good mechanical strength at sintering temperature of 90$0^{\circ}C$. As the holding time increased, neck between the particles grew progressively, but shrinkage of the specimen did not occur, implying that the porosity remained constant during the whole sintering process.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.15 no.1
• /
• pp.310-317
• /
• 1995

Crack detection technique by ultrasonics in structures and components made of austenitic stainless steel often loses its reliability due to the material characteristics during inservice inspection of nuclear power plants, especially in the area of detection and sizing in centrifugally cast stainless steel pipings. In order to understand and overcome this problem, computer program for tracing the ultrasonic rays within material has been developed to simulate the process of defect detection within weld. The program simulates through transmission and reflection technique in crack detection of austenitic stainless steel as well as ultrasonic beam propagation through multiple media including stainless steel cladding interface.

• Transactions of the Korean hydrogen and new energy society
• /
• v.15 no.4
• /
• pp.266-273
• /
• 2004

The objective of this study is an understanding of stress corrosion cracking of metals that is recognized to mostly limit the lifetime of the structural materials by comparing the features of delayed hydride cracking of zirconium alloys with those of stress corrosion cracking (SCC) of Ni-based alloys and hydrogen cracking of stainless steels. To this end, we investigated a dependence of delayed hydride cracking (DHC) velocity on the applied stress intensity factor and yield strength, and correlated a temperature dependence of the striation spacing and the DHC velocity. We reviewed a similarity of the features between the DHC of zirconium alloys, the SCC of Ni-based alloys and turbine rotor steels, and the hydrogen cracking of stainless steels and discussed the SCC phenomenon in metals with our DHC mode.

• Nuclear Engineering and Technology
• /
• v.26 no.1
• /
• pp.140-148
• /
• 1994

A qualification test was performed for the iron removal chemical cleaning of the secondary side of nuclear steam generators at the selected temperature, 1$25^{\circ}C$, higher than the standard application temperature, 93$^{\circ}C$. The field cleaning condition for a nuclear unit was tested in a bench scale test loop including a SUS 316 stainless steel autoclave with one gallon capacity as a test vessel. The kinetics of sludge dissolution, corrosion of the secondary side materials and change of solvent chemistry were monitored. Test results indicated that more thorough cleaning was accomplished in less than half of the cleaning time required at 93$^{\circ}C$. And the total corrosions of the secondary side materials were found to be less than the values at 93$^{\circ}C$. While the solvent is recirculated and heated by an external chemical cleaning equipment for the conventional 93$^{\circ}C$ process, the secondary side is heated by the lateral heat of the primary coolant without the recirculation of the cleaning solution, and the solvent is mixed by vigorous boiling induced by periodic ventilation for the high temperature process. The requirement that the reactor coolant pumps should be running during the cleaning operation is the major disadvantage of the high temperature process which also should be considered when chemical cleaning is planned for steam generators under operation.