• 한국산업안전학회:학술대회논문집
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• 한국안전학회 2003년도 추계 학술논문발표회 논문집
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• pp.38-43
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• 2003

증기발생기 전열관의 마멸은 유체 유발 진동(flow induced vibration)에 의한 전열관과 증기발생기 상부 지지구조물 사이에서 발생하게 되며 원통 지지대(stay cylinder)상부의 중앙 공공(central cavity) 주변에 집중적으로 발생되는 것으로 보고되고 있다. 국내에서는 1997년 영광 4호기의 증기발생기에 마멸 손상이 보고된 이후 영광 3호기와 울진 3, 4호기에서도 마멸 손상이 발견되고 있으며, 외국에서는 1992-1993년 기간동안 대략 500∼600 개의 전열관이 마멸에 의해 관막음(p1u99ing)된 것으로 보고되었다.(중략)

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2002년도 추계학술대회논문집
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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.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2002년도 추계학술대회논문초록집
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• pp.390.1-390
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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 fur 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 mode shape. We also investigated the turbulent excitation effect.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2003년도 춘계학술대회논문집
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• pp.670-676
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• 2003

To investigate the group plugging effect the fluid-elastic instability analysis has been performed on various column and row number of the KSNP steam generator lutes. This study compares the stability ratio of the plugged tube with that of the intact one. The information on the thermal-hydraulic data of the steam generator have been obtained by using the ATHOS3-MOD1 code with and without the thermal energy transfer at the plugged region. Both of the boundary conditions of hot-leg temperature and feedwater mass flow rate are fixed for this investigation. From the results of this study the stability ratios inside the group plugging zone are decreased slightly. At the outside of group plugging zone, however, most of the stability ratios tend to be increased.

• 한국압력기기공학회 논문집
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• 제16권2호
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• pp.10-17
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• 2020

Securing the integrity of steam generator tubes is an essential requirement for safe operation of nuclear power plants. Therefore, tubes that do not satisfy integrity requirements are no longer usable and must be repaired according to the related requirements. In general, the repair criterion is that the damage depth is more than 40% of the tube wall thickness. However, the plugging limit can be changed and be applied, provided a technical proof is given that integrity can be secured against specific degradation at a specific plants and that approval can be obtained from a regulatory agency. A typical example is alternative repair criteria for defects within the tube sheet or tube support plates. In this paper, a background of establishing the plugging limit for steam generator tubes and changes in maintenance criteria are reviewed as examples.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 추계학술대회논문집B
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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.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2006년 추계학술발표대회 개요집
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• pp.38-40
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• 2006

Nuclear power plants are urged to reduce operating and maintaining costs to remain competitive as well as to increase the safety preventing the radioactive material to the atmosphere. To reduce the cost and to increase the safety, the inspection of balance-of-plant heat exchanger becomes important. However, there are some problems for plugging the heat exchanger tubes since the criterion and its basis are not clearly described. The codes and standards related to show the tube plugging criteria may not exist currently. In this paper, a method to establish the tube plugging criteria of BOP heat exchangers is introduced and the tube plugging criteria for the non-regenerative heat exchanger in the steam generator blow-down system of nuclear power plant. This method relies on the similar method used to establish the plugging criteria for the steam generator tubes.

• 한국압력기기공학회 논문집
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• 제11권1호
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• pp.25-32
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• 2015

Many crack flaws have been detected at tubing seal weld zone in Steam Generator with Alloy600 material tubes. it is very difficult to discriminate the signal from parent material and that from weld zone because weld zone gives complicated ECT signal due to manufacturing processes. ECT signal analysis in the seal weld zone is very important since the tubes with crack in this area shall be out of service by plugging. The purpose of this paper is to distinguish between Parent material and Weld Zone ECT signal in the seal weld zone through the demonstration test of Steam Generator seal weld mock-up.

• 대한기계학회논문집A
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• 제20권9호
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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.