Papers by Keyword: Alloy 690

Abstract: Aluminides were formed on Ni-Cr-Fe based superalloy 690 substrates using pack aluminization process at 1273 K in controlled atmosphere. Thermal oxidation of aluminized specimens was carried out at 1273 K for a total period of 4 hours in air. The thermally grown oxide layer was examined using X-ray diffraction (XRD) studies on top surface and scanning electron microscopy (SEM) with energy dispersive spectroscopy (EDS) analysis along the cross-section of the sample. The oxide layer developed on aluminized superalloy 690 substrate consisted of Al₂O₃ layer with a thickness of about 2 μm. The oxidized specimens were exposed in nitrate-based environment (simulated high-level nuclear liquid waste) at 373 K for a total period of 216 hours. A good adherence of aluminide coatings was noticed even after prolonged exposure in nitrate-based solution with a little amount of material dissolution from the edges of the specimens. XRD studies on exposed specimen indicated existence of Al₂O₃ layer on the top surface, which is believed to have resulted in good adherence of aluminide coatings.

Abstract: The occurrence of microcracks, especially ductility-dip crack in multipass weld metal during GTAW and laser overlay welding processes of Ni-base alloy 690 was predicted by the mechanical approach. The stress/strain analysis in multipass welds was conducted using the thermo elasto-plastic finite element method. The brittle temperature range for ductility-dip cracking (DTR) of the reheated weld metal was determined by the Varestraint test. Plastic strain in the weld metal accumulated with applying the weld thermal cycle in multipass welding. The plastic strain-temperature curve in the La free weld metal did not cross the DTR in the cooling stage of GTAW process, however, it crossed the DTR in the cooling stage of reheating process by subsequent welding. On the other hand, the plastic strain-temperature curves of any weld passes in the La added weld metal did not cross the DTR. Ductility-dip cracks occurred in the La free weld metal except for the final layer, however, any ductility-dip cracks did not occur in the La added weld metal during multipass welding. It could be understood that ductility-dip crack would occur during not only single-pass welding but also multipass welding when plastic strain intersected the DTR.

Abstract: The stress corrosion cracking (SCC) susceptibility of Alloy 600 MA, Alloy 600 TT, Alloy 800, and Alloy 690 TT were investigated in a deaerated 0.01 M solution of sodium tetrathionate using reverse u-bend test samples at 340 °C. The results showed that SCC occurred in all alloys, excluding Alloy 690 TT. The SCC susceptibility of the alloys increased in the following order: Alloy 690 TT, Alloy 800, Alloy 600 TT, and Alloy 600 MA. The SCC susceptibility decreased with an increase in the chromium content of the alloys. The results of the deposits and spectra taken from an energy dispersive X-ray system confirmed the existence of a reduced sulfur causing SCC.

Abstract: Several chemicals were studied to suppress the damage due to stress corrosion cracking (SCC) of steam generator (SG) tubes in nuclear power plants. The polarization curves showed that the electrochemical properties on the surface of Alloy 600 MA changed with the addition of inhibitors. The SCC tests were conducted by using a m-RUB specimen in a 10% NaOH solution at a temperature of 315°C. The effects on the SCC of the compounds, TiO2, TyzorLA and CeB6, were tested for several types of SG tubing materials. The test with the addition of TiO2 (P25) and CeB6 showed an effect in decreasing the SCC for the SG tubing material. However, CeB6 caused some more SCC for Alloy 800. The penetration property into a crevice of the inhibitors was investigated by using Alloy 600 specimens with different gap sizes and an AES analysis was performed on the oxide layer of the specimen.

Abstract: Fretting wear generated by flow induced vibration is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Understanding of tube wear characteristics is very important to keep the integrity of the steam generator tubes to secure the safety of the nuclear power plants. Experimental examination has been performed for the purpose of investigating the impact fretting. Test material is alloy 690 tube and 409 stainless steel tube supports. From the results of experiments, wear scar progression is investigated in the case of impact-fretting wear test of steam generator tubes under plant operating conditions such as pressure of 15MPa, high temperature of 290
C and low dissolved oxygen. Hammer imprint that is actual damaged wear pattern, has been observed on the worn surface. From investigation of wear scar pattern, wear mechanism was initially the delamination wear due to cracking the hard oxide film and finally transferred to the stable impact-fretting pattern.

Abstract: Intergranular attack/stress corrosion cracking of Alloy 600 continues to be an issue in the tube/tube support plate crevices and top of tubesheet locations of recirculating steam generators and in the upper bundle of free span superheated regions of once through steam generators (OTSG). Recent examinations of degraded pulled tubes from several plants suggest possible lead involvement in the degradation. Laboratory investigations have been performed to determine the factors influencing lead cracking in Alloy 600 and Alloy 690 steam generator tubes. The test environment is believed to be prototypical, with the addition of lead oxide, of a concentrated liquid phase existing in the pores of thin deposits on upper bundle tubes of an OTSG. Highly strained reverse U-bend specimens were tested at controlled electrochemical potentials. Maximum susceptibility was at open circuit potential, unlike cracking of Alloy 600 in caustic and acid sulfate environments where maximum susceptibility occurs when specimens are polarized above the open circuit potential. Transgranular, intergranular and mixed mode cracking was observed and in all Alloy 600 conditions tested (mill annealed, sensitized, thermally treated) while thermally treated Alloy 690 has so far resisted cracking. A film rupture/anodic dissolution model with displacement plating of Pb preceding passive film formation is consistent with the experimental observations