• Transactions of the Korean Society for Noise and Vibration Engineering
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• 제14권4호
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• pp.327-335
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• 2004

This study investigates the safety assessment of the potential for fretting-wear damages caused by foreign object in operating nuclear power plants. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for the helical type tubes with various conditions. The wear rate of helical type tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the vibration of the tube on the remaining life of the tube. In addition, addressed is the effect of the external pressure on the vibration and fretting-wear characteristics of the tube.

• Journal of Ocean Engineering and Technology
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• 제21권2호
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• pp.50-59
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• 2007

Steam generator tubes are degraded from wear, stress corrosion cracking, rupture and fatigue and so on. Therefore, the failure assessment of steam generator tube is very important for the integrity of energy plants. In the steam generator tubes, sometimes, the local wall thinning may result from severe degradations such as erosion-corrosion damage and wear due to vibration. In this paper, the elasto-plastic analysis was performed by FE code ANSYS on steam generator tubes with wall thinning. Also, the four-point bending tests were performed on the wall thinned specimens, and then it was compared with the analysis results. We evaluated the failure mode, fracture strength and fracture behavior from the experiment and FE analysis. Also, it was possible to predict the crack initiation point by estimating true fracture ductility under multi-axial stress conditions at the center of the thinned area from FE analysis.

• Proceedings of the KSME Conference
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• 대한기계학회 2007년도 춘계학술대회A
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• pp.250-255
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• 2007

Steam generator tubes play an important role in safety because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the nuclear power plant. For this reason, the integrity of the tubes is essential in minimizing the leakage possibility of radioactive water. The integrity of the tubes is evaluated based on NDE (non-destructive evaluation) inspection results. Especially ECT (eddy current test) method is usually used for detecting the flaws in steam generator tubes. However, detection capacity of the NDE is not perfect and all of the "real flaws" which actually existing in steam generator tunes is not known by NDE results. Therefore reliability of NDE system is one of the essential parts in assessing the integrity of steam generators. In this study POD (probability of detection) of ECT system for ODSCC in steam generator tubes is evaluated using multivariate logistic regression. The cracked tube specimens are made using the withdrawn steam generator tubes. Therefore the cracks are not artificial but real. Using the multivariate logistic regression method, continuous POD surfaces are evaluated from hit (detection) and miss (no detection) binary data obtained from destructive and non-destructive evaluation of the cracked tubes. Length and depth of cracks are considered in multivariate logistic regression and their effects on detection capacity are evaluated.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• 제16권1호
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• pp.49-55
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• 2020

Creep rupture data for Alloy 690 steam generator tubes in a pressurized water reactor are essentially needed to demonstrate a severe accident scenario on thermally-induced tube failures caused by hot gases in a damaged reactor core. The rupture data were obtained using the tube specimens under different applied-stress levels at 650℃, 700℃, 750℃, 800℃, and 850℃. Important creep constants were proposed using various creep laws in terms of Norton power law, Monkman-Grant (M-G) relation, damage tolerance factor (λ), and Zener-Hollomon parameter (Z). In addition, a creep activation energy (Q) value for Alloy 690 tube was reasonably determined using experimental data. Creep behaviors such as creep strength, creep rates, rupture elongation showed the results of temperature dependence well. Modified M-G plot improved a correlation of the creep rate and rupture life. Damage tolerance factor for Alloy 690 tubes was found to be λ =2.20 in an average value. Creep activation energy for Alloy 690 tube was optimized for Q=350 (kJ/mol). A plot of Z parameter obeyed a good linearity, and the same creep mechanism was inferred to be operative in the present test conditions.

• Proceedings of the KSME Conference
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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.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• 제15권1호
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• pp.19-24
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• 2003

The objectives of this paper are to develop a new method of wettability mea-surement and to study the effect of surface roughness on the wettability in a $H_2O$/LiBr falling film absorber. Two absorber tubes with micro-scale roughness and a bare tube are tested in a falling film absorber installed in a test rig. Inlet solution temperature, concentration and mass flow rate are considered as key parameters. A new method is proposed to estimate the wettability of a tube by measuring a minimum mass flow rate to wet the tube completely. The wettability for the structured surfaces was higher than that for the bare tube. The wettability decreased linearly along the vertical location. The wettability increased with increasing the solution temperature and the solution mass flow rate. The experimental correlations of the wettability for the bare and the micro-hatched tubes were developed with error bands of$\pm20%\;and\;\pm10%$, respectively. This work can be used in the design of absorbers with micro-scale roughness.

• Proceedings of the SAREK Conference
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• 대한설비공학회 2009년도 하계학술발표대회 논문집
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• pp.266-271
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• 2009

The present study numerically investigated the effect of the geometry of the flattened tube on the thermal performance of a high temperature generator (HTG) of a double effect LiBr-water absorption system. The heat transfer tubes of the HTG were arranged behind a metal fiber burner. The heat transfer of the tubes of HTG were consisted with a set of circular and flattened tubes in series. FLUENT, as a commercial code, was applied for estimating the thermal performance of the HTG. Key parameters were the tube arrangement in the HTG. Temperature and velocity profiles in the HTG were calculated to estimate the thermal performance of the HTG. The heat transfer rate of a HTG tube was increased, and the gas temperature around the flattened tube was decreased as the pitch ratio was increased. The heat transfer rate for the circular tube bundle with the pitch ratio of 2.48 were larger by 10% respectively than that of 2.10 and the heat transfer rate for the flattened tube bundle with the pitch ratio of 1.88 were larger by 36% respectively than that of 1.63.

• Journal of the Korea Academia-Industrial cooperation Society
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• 제14권4호
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• pp.2043-2048
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• 2013

OPR-1000 CE Steam Generator, of which tube material is composed of Alloy-600 HTMA in nuclear power plant, secondary side is generated ODSCC(Outside Diameter Stress Corrosion Cracking) due to the accumulated sludge. ODSCC is centered around the tube sheet and is being affected depending on the height of the sludge. Chemical cleaning was carried out for a top-of-the-tube sheet(TTS) of Steam Generator in order to decrease corrosive condition of the secondary side of Steam Generator tubes and suppress the occurrence of stress corrosion cracking. The amount of sludge removal was 259.2kg. The height of the accumulated sludge was reduced from 0.71 to 0.34 inches. Corrosion rate as the maximum 2.34 mils was satisfied to within EPRI (Electric Power Research Institute) recommendation(10 mils).

• Transactions of the Korean Society of Pressure Vessels and Piping
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• 제7권1호
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• pp.35-40
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• 2011

If a plugged tube in a steam generator is broken, it may damage nearby intact tubes. To prevent this damage, it is recommended that a stabilizer is installed into the plugged tube. However, the installation cost of a stabilizer is very high. So studies are required to determine the conditions on which the installation is necessary. For this purpose static loads and dynamic loads applied on a tube should be known to estimate the residual strength and remaining fatigue and wear life of a plugged tube. Two-dimensional and three-dimensional computational fluid dynamics (CFD) analyses are performed to obtain the drag coefficient for cross flow to a tube. Using the obtained drag coefficient, the static load can be estimated and the residual strength of a plugged tube can be calculated. An inclined flow problem is also analyzed and the vertical and horizontal forces are obtained and discussed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• 제7권4호
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• pp.51-56
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• 2011

Steam generator is one of the main equipments that affect safety and long term operation in nuclear power plants. Fluid flows inside and outside of the steam generator tubes and induces vibration. To prevent the vibration the tubes are supported by AVB (anti vibration bar). When the steam generator tube contact to AVB, it is damaged by the accumulation of wear and corrosion. Therefore studies are required to determine the effects of the gap between the steam generator tube and AVB. In order to obtain the stress and the displacement distributions of the steam generator tube, three dimensional finite element analyses were performed by using the commercial program ANSYS. Using the calculated the stress and the displacement distributions, the static residual strength of the steam generator tube can be evaluated. The results show that the stress and displacement of the steam generator tube increase significantly compared with the results from a zero-gap model.