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

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

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

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

• Tribology and Lubricants
• /
• v.17 no.5
• /
• pp.380-385
• /
• 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
• /
• v.17 no.10
• /
• pp.555-559
• /
• 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.

• Proceedings of the Korean Society of Laser Processing Conference
• /
• 1999.11a
• /
• pp.39-42
• /
• 1999

• Korean Journal of Materials Research
• /
• v.14 no.4
• /
• pp.251-259
• /
• 2004

In the development of the advanced spent fuel management process based on the molten salt technology, it is essential to choose the optimum material for the process equipment handling molten salt. Corrosion behavior of Ni 200 and Ni-base alloys in molten salt of LiCl-$Li_2$O under oxidation atmosphere was investigated in the temperature range of $650～800^{\circ}C$ for 24～312 hrs. The order of corrosion rate was Ni 200 > Inconel 690 > Inconel 601 > Inconel 600. Inconel 600 alloy showed the highest corrosion resistance among the examined alloys, but Ni 200 exhibited the highest corrosion rate. Corrosion products of Inconel 600 and Inconel 601 were $Cr_2$$O_3$ and $NiFe_2$$O_4$. In case of Inconel 690, a single layer of $CrO_2$$O_3$ was formed in the early stage of corrosion and an outer layer of $NiFe_2$O$_4$ and inner layer of $Cr_2$$O_3$ were formed with increase of corrosion time. Inconel 600 showed local corrosion behavior and Inconel 601, 690 showed uniform corrosion behavior.

• Tribology and Lubricants
• /
• v.21 no.1
• /
• pp.33-38
• /
• 2005

Fretting is the oscillatory motion with very small amplitudes, which usually occurs between two solid surfaces in contact. Fretting wear is the removal of material from contacting surfaces through fretting action. Fretting wear of steam generator tubes in nuclear power plant becomes a serious problem in recent years. The materials for the tubes usually are Inconel 690 (I-690) and Inconel 600 (I-600). In this paper, fretting wear tests for I-690 and I-600 were performed under various applied loads in water at room temperature. Results showed that the fretting wear loss of I-690 and I-600 tubes was largely influenced by stick-slip. The fretting wear mechanisms were the abrasive wear in slip regime and the delamination wear in stick regime. Also, I-690 had somewhat better wear resistance than I-600.

• Journal of Welding and Joining
• /
• v.17 no.1
• /
• pp.82-87
• /
• 1999

The basic remote sleeve repair-welding technology by the pulsed Nd:YAG laser for increasing the lifetime of the steam generator tube in a nuclear power plant has been developed. The relationship between the connection width and welding parameters have been investigated for the fundamental research to apply the sleeve-repair-welding technique to the nuclear industry. The Inconel 600 tube and Inconel 690 sleeve used for high temperature and high pressure service were welded as round lap welding by Nd:YAG laser. It was observed that the tensile shear strength, 340MPa of the welded specimen is equivalent to about 60% of that of the base metal (Inconel 600), 550MPa. The difference between the hardness of the base metal and that of the laser welds was about 10%. Ductile fracture was partly occurred in the weld but the cracking has not been observed. In spite of absence of the crack, the strength of welds was not sufficient in terms of the tensile shear strength.