• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2002.11b
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• pp.921-926
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• 2002

In this study, we investigate the plugging effect on the CE type steam generator tube. The natural frequency and mode shape will be changed due to decrease of the effective mass distribution along the tube. We compared the variation of stability ratio for plugged tube with that for unplugged one. The natural frequency increased because of removing the cooling water inside the steam generator tube, but the stability ratio decreased inversely because of changing the vibrational model shape. We also investigated the turbulent excitation effect.

• Geotechnical Engineering
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• v.11 no.4
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• pp.25-36
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• 1995

A specially designed model open -ended pile, which was composed of inner tube and outer tube, wry driven in the pressure chamber by two diffenent intallation methods, that is, impact -driving and vibratory driving, and static compression loading test was done for that pile. Through the measurement of bearing capacities in the separated resisting parts of open -ended pile, bearing mechanism of open-ended pile and soil plugging behaviors for different installation methods were studied. It appears that 20% out of soil plugging force of impact -driven pile was developed during driving, while the rest was developed during static compression loading and t.he magnitude of confining pressure applied to the chamber did not affect soil plugging behavior. Also. it appears that, soil plugging force of vibratory pile was not developed during driving, while it was developed weakly as about 0.5~0.7 times as that of impact pile during static compression loading. and the confining pressure of pressure rhamber had an effect on the soil plugging. In the ultimate loading condition unit soil plugging force did not approach to the failure condition.

• Journal of Environmental Science International
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• v.7 no.5
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• pp.669-677
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• 1998

A porous $\alpha$-alumina tube of 2.5 mm O.D. and 1.9 mm I.D. was used as the support of an inorganic membrane. Macropores of the tube, about 150 nm in size, were plugged with silica formed by thermal decomposition of tetraethylorthosilicate at $600^{\circ}C$. The forced cross-flow CVD method that reactant was evacuated through the porous wall of the support was very effective in plugging macropores. The H$_2$ permeance of the prepared membrane was of the order of $10^{-8}/ molㆍs^{-1}/ㆍm^{-2}/. Pa{-1}$/, while the $N_2$ permeance was below $10^{-11}/ molㆍs^{-1}/ㆍm^{-2}/ㆍPa^{-1}$/ at $600^{\circ}C$. This was comparable to that of silica-modified Vycor glass whose size was 4 nm.

• Proceedings of the KSME Conference
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• 2001.11b
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• pp.859-866
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• 2001

During the Pre-service Inspection for Wolsong Unit 3&4 in 1997/1998 respectively, 17 Distorted Roll Transition indications(over expanded beyond tubesheet secondary face) were identified at the Unit 4 (S/G B, D). Six(6) tubes out of these tubes were plugged in 1998. However the first Periodic Inspection identified additional 110 indications in 1999 and 2000. The additionally identified 110 indication call, not reported at the Pre-service Inspection, are; 2 Not-Finally-Expanded-Tubes and 108 Distorted Roll Transition tubes. Design limit of each Steam Generator tube Plugging is 6.4%. Plugging was performed by the Steam Generator manufacturer under the warranty. When Distorted Roll Transition indications were first identified on the Unit 4 in 1998 the degree of Over-expansion was measured using an inner dial-gage to make the disposition of Nonconformance report. 2 Not-Finally-Expanded-Tubes were plugged and 10 tubes out of 108 Distorted Roll Transition Tubes were also plugged as a preventive measure.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.351.1-351
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• 2002

• Transactions of the Korean Society of Mechanical Engineers A
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• v.20 no.9
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• pp.2894-2900
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• 1996

The stream generator tubes represent an integral part of a major barrier against the fission product release to the environment. So, the rupture of these tubes could permit flow of reactor coolant into the secondary system and injure the safety of reactor coolant system. Therefore, if the crack was detected during In-Service Inspection of tubes the cracked tube should be evaluated by the pulgging criteria and plugged or not. In this study, the fracture mechanics evaluation is carried out on the thru-wall axial crack due to Primary Water Stress Corrosion Cracking in the roll transition aone of steam generator tube to help the assurence the integrity of tubes and estabilish the plugging criteria. Due to the Inconel which is used as tube material is more ductile than others, the plastic instability repture theory was used to calculate the critical and allowable crack length. Based on Leak Before Break concept the leak rate for the critical crack length and the allowable leak rate are compared and the safety of tubes was given.

• Proceedings of the KSME Conference
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• 2003.11a
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• pp.55-60
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• 2003

As operating time of heat exchangers progresses, fouling generated by water-borne deposits increases and thermal performance decreases. The fouling is known to interfere with normal flow characteristics and reduce thermal efficiencies of heat exchangers. The heat exchangers of nuclear power plants have been analyzed in terms of the heat flux and heat transfer coefficient at test conditions based on the ASME OM-S/G-Part 2 as a means of heat exchanger management. It is hard to estimate the heat performance trend and to establish the future management plan. This paper describes the fouling evaluation method which can evaluate the thermal performance for heat exchangers and estimate the future fouling variations and the plugging margin evaluation method which can reflect the current fouling level developed in this study. To develop the fouling and plugging margin evaluation methods for heat exchangers, fouling factor was introduced based on the ASME O&M codes and TEMA standards. For the purpose of verifying the two evaluation methods, the fouling and plugging margin evaluations were performed for a component cooling heat exchanger in a nuclear power plant.

• Applied Chemistry for Engineering
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• v.9 no.5
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• pp.674-679
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• 1998

Small water leak experiment was carried out in liquid sodium atmosphere using a specimen of ferrite steel, which will be expected to be a material of the heat transfer tube of liquid metal fast breeder reactor. Self-plugging phenomena of leak path could be explained by the products of reaction and corrosion by sodium-water reaction. Also, re-opening mechanism of self-plugged path could be explained by the thermal transient and vibration of heat transfer tube. As a result, perfect re-opening time of self-plugged leak path was observed to be 129 minutes after water leak initiation. Re-opening shape of a specimen was appeared with double layer of circular type, and re-opening size of this specimen surface was about 2 mm diameter on sodium side.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.462.1-462.1
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.131.2-131
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• 1999