• Proceedings of the KSME Conference
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• 2003.04a
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• pp.49-54
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• 2003

Most of the operating pressurized water reactors (PWRs)has chosen Inconel 600 as steam generator tubing. The long-term operation of steam generators showed that the use of this material induced localized corrosion damages. The current trend is using Inconel 690 as a tube material for the replacement steam generators. Based on the current trend, we have chosen Inconel 690 for the advanced Power Reactor 1400 (APR1400) steam generator tube material. In this paper, we examined the technical consideration in this modification: the effect of chemical composition, thermal conductivity, corrosion resistance and wear characteristics

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.103-116
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• 2006

By executing corrosion experiment on Inconel 600, 690 used to material of S/G tube in domestic NPP, this paper show estimation of amount of product such as Co-58, Co-60, Cr-51, Mn-54, Fe-59 which are main exposure cause to the workers in NPP. Therefore, Making the 12 samples consisted of Inconel 600, 690, whole corrosion experiment was carried out for 60 days(each pH by 20 days). The conditions of those tests were similar or more harsh than actual conditions of domestic NPP. The Glow Discharge Spectrometer(GDS) was used for quantitative analysis of results. The results of using GDS, the Inconel 600 corrodes more than Inconel 690 at pH 7 and pH 9. However, it is observed that Inconel 690 corrodes more than Inconel 600 at pH 4. Those results is estimated that test sections had the effect of transient. The long terms of experiment is required to minimize and solve the problem.

• Journal of Welding and Joining
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• v.20 no.2
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• pp.82-89
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• 2002

In order to clarify hot cracking phenomena occurred in Inconel 690 welds and it's prevention, in this study, the cracking behavior and the influence of welding variables on cracking in Inconel 690 overlay welds on Mn-Ni-Cr-Mo steel(SA 508 cl.3) for pressure vessel were investigated by using mock-up test. The main results are as follows: The cracks in Inconel 690 overlay welds were mainly generated near the start and the end part of welding beads adjacent to STS 309L welded outside of Inconel 690 welds. Most of the cracks showed typical solidification crack, and also it was assumed that there was possibility of liquation cracking in HAZ. The existence of Nb constituents or concentration of Nb was recognized on the fracture facets of the solidification cracks in the welds by SMAW. Therefore Nb was considered to be the main factor of the solidification cracking. As the weld heat input was more increased and the weld bead length was longer, the extent of cracking was more increased. Moreover the extent of cracking was considerably decreased by changing of welding sequence to the start and the end part of welds. Hot cracking in welds by GTAW was considerably decreased as compared with that of SMAW. And cracks were well generated in the Inconel 690 overlay welds adjacent to 575 309L welds. This means that the hot cracking susceptibility of Inconel 690 welds was largely varied by chemical components and/or compositions of filter metals, base metals and neighboring welds.

• Journal of Welding and Joining
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• v.20 no.3
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• pp.74-83
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• 2002

In order to investigate the hot cracking phenomena of the Inconel 690 overlay welds, the hot cracking test(modified Varestraint test) was performed by varying augmented strain and welding speed in GTAW. A total of five kinds of specimens was used in the hot cracking test. The specimens used were two kinds of one layer specimens that were composed of just Inconel 690 deposited metal and three kinds of double layer specimens that were composed with as upper part of Inconel 690 weld metal and lower part of SA508 cl.3 or STS 309L. The main results are as fo11ows: In the welds composed of just Inconel 690, as the augmented strain was more increased and the welding speed was more decreased, the extent of cracking was more increased. And these cracks were mainly solidification cracks, and liquation cracks were also observed partially in HAZ. And hot cracking susceptibily of Inconel 690 welds by using filler metals containing Nb were higher than that of the welds with Nb free. The hot cracking susceptibility of Inconel 690 weld metal was increased with dilution of SA 508 cl.3 and STS 309L, and the influence of SA 508 cl.3 was higher than that of 575 309L. The results of calculation of $TSC\Delta$ and $TLC\Delta$ in Inconel 690 weld metal changed by dilution with base metal or neighboring welds were agreed well with the results of hot cracking test in this study. Therefore, it was expected that the hot cracking was considerably decreased or prevented by using proper welding conditions such as lower heat input, filler metals, base metal and neighboring welds.

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.110.2-110.2
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• 2005

• Proceedings of the Materials Research Society of Korea Conference
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• 2003.11a
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• pp.91-91
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• 2003

Sliding wear behaviors of Inconel 600 and 690 were investigated at room temperature in air. In the present study, Archard's equation which has low reliability was modified. In the prediction of wear volume by Archard's equation, the reliabilities of Inconel 600 and 690 were about from 26.3% to 45.7% and from 69. l% to 88.6%, respectively, The sliding wear behaviors of Inconel 600 and 690 turned out to be influenced by their stacking fault energy, and the fact was confirmed by using TEM and micro-hardness test Based on experimental results, the wear coefficient was modified as a function of the sliding distance. The calculation with the modified wear equation showed that the reliability of Inconel 600 tested with 409 ferritic stainless steel increased from 45.7% to 93.4%.

• Tribology and Lubricants
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• v.17 no.5
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• pp.380-385
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• 2001

In nuclear power steam generators, high flow rates can induce vibration of the tubes resulting in fretting wear damage due to contacts between the tubes and their supports. In this paper the fretting wear tests and the sliding wear tests were performed using the steam generator tube materials of Inconel 600 and 690 against STS 304. Sliding tests with the pin-on-disk type tribometer were done under various applied loads and sliding speeds at air environment. Fretting tests were done under various vibrating amplitudes and applied normal loads. From the results of sliding and fretting wear tests, the wear of Inconel 600 and 690 can be predictable using the work rate model. Depending on normal loads and vibrating amplitudes, distinctively different wear mechanisms and often drastically different wear rates can occur. It was found the results that the wear coefficients for Inconel 600 and 690 were 262.3$\times$10$\^$-15/Pa$\^$-1/ and 209.2$\times$10$\^$-15/Pa$\^$-1/, respectively. This study shows that Inconel 690 can provide much better wear resistance than Inconel 600 in air.

• Korean Journal of Materials Research
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• v.17 no.10
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• pp.555-559
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• 2007

The mechanical properties and microstructure of Inconel 690 and 600 alloys with various cooling rates were investigated. Optical microscopy and scanning electron microscopy observations indicated that in case of the cooling rate of $0.5^{\circ}C/min$, discontinuous carbides along the grain boundaries were formed and when the cooling rate was $10^{\circ}C/min$, continuous carbides were formed in Inconel 690 and 600 alloys. For the annealed Inconel 690 alloy with high Cr content, a lot of annealing twins, which led the preferential growth of (111) planes, were observed. However, the annealed Inconel 600 alloy with low Cr content showed a few annealing twins and the preferential growth of (200) planes. Inconel 600 alloy had a larger value of ultimate tensile strength (UTS) than Inconel 690 alloy.

• Transactions of the Korean Society of Machine Tool Engineers
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• v.11 no.4
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• pp.54-61
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• 2002

Inconel 690 alloy has widely been used in power plant and high temperature facilities because it has high thermal resistance and toughness. But we have little design data about the creep behaviors of the alloy. Therefore, in this study, an apparatus has been designed and built for conducting creep tests under constant load conditions. A series of creep tests on Inconel 690 alloy have been performed to get the basic design data and life prediction of inconel products and we have gotten the following results. First, the stress exponents decrease as the test temperatures increase. Secondly, the creep activation energy gradually decreases as the stresses become bigger. thirdly, the constant of Larson-Miller Parameters on this alloy is estimated about 10. And last the fractographs at the creep rupture show both the ductile and the brittle fracture according to the creep conditions.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.3
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• pp.432-439
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• 2003

Inconel 690 for nuclear steam generator tube has more Chromium than the conventionally used Inconel 600 in order to increase the corrosion resistance. To evaluate the tribological characteristics of Inconel 690 under fretting condition the fretting tests were carried out in air and elevated temperature water. Fretting tests of the cross-cylinder type were done under various vibrating amplitudes and applied normal loads in order to measure the friction forces and wear volumes. From the results of fretting wear tests. the wear of Inconel 690 can be predictable using the work rate model. The amounts of friction forces were proportional to relative movement between two fretting surfaces. The friction coefficients were decreased as increasing the normal loads and deceasing the vibrating amplitudes. Depending on fretting environment, distinctively different wear mechanisms and often drastically different wear rates can occur It was found that the fretting wearfactors in air and water at 2$0^{\circ}C$, 5$0^{\circ}C$, and 8$0^{\circ}C$ were 7.38 $\times$ $10^{-13}$$Pa^{-1}$, 2.12 $\times$$10^{-13}$$Pa^{-1}$, 3.34$\times$$10^{-13}$$Pa^{-1}$and 5.21$\times$$10^{-13}$$Pa^{-1}$, respectively flexibility to model response data with multiple local extreme. In this study, metamodeling techniques are adopted to carry out the shape optimization of a funnel of Cathode Ray Tube, which finds the shape minimizing the local maximum principal stress. Optimum designs using two metamodels are compared and proper metamodel is recommended based on this research.