Nuclear Engineering Division

Corrosion and Mechanics of Materials

Light Water Reactors

 

Steam Generator Tube Integrity Program

Steam generator tubes, which account for more than 50% of the primary pressure boundary surface of PWRs, have experienced in-service corrosive and mechanical degradation of various forms since the beginning of PWR commercial operation in the late 1950s. Various forms of degradation have resulted in the plugging of well over 100,000 tubes to date around the world. In addition, 68 steam generators in 22 U.S. plants had been replaced by the end of 1998 at a cost of about $100 to $200 million each, and more replacements are underway or planned. Environmentally induced degradation through intergranular SCC and intergranular attack is the most serious degradation process at present. This degradation commonly occurs in crevice regions at tube support plate and tube sheet locations or under sludge piles, although intergranular SCC has also been observed in the free span of the tubes. Because of its variable and often complex morphology, this cracking can be difficult to detect and size by conventional inspection techniques, and the failure pressure and leak-rate behaviors of degraded tubes are not readily predictable.

One of the objectives of this NE program is to evaluate and experimentally validate models to predict potential degradation modes and provide guidance for operational assessments. Model development requires better understanding of crevice conditions and SCC initiation, evolution, and growth. The models will be validated by comparing the predictions from field experience with respect to SCC of Alloy 600 steam generator tubes. The methodology benchmarked by Alloy 600 field experience will then be used to predict the behavior in the field of Alloy 690 steam generator tubes based on the laboratory data for Alloy 600 and 690 and the field experience for Alloy 690.

The models for predicting ligament rupture, unstable burst, and leak rate of flawed Alloy 600 tubes developed in the program have been incorporated into CANTIA, an integrity assessment code for steam generator tubes that was originally developed by Dominion Engineering for the Canadian Nuclear Safety Commission. Several other modifications were also made in Argonne/CANTIA, which can model either axial or circumferential cracks, but not both simultaneously. An integrated mechanistically based model has also been developed (under a subcontract with Dr. Roger W. Staehle) to predict the secondary-side SCC failure of steam generator tubes under normal operating conditions. This effort is continuing with development of a quantitative model that is physically based on the prediction of secondary-side SCC initiation for the different submodes of SCC in given environments.

An autoclave facility for corrosion and stress corrosion initiation tests has been constructed. The facility consists of two independent recirculating loops. Each loop has an 8-L (2.1 gal) Hastelloy C276 autoclave vessel, high pressure pump, furnace, feed water system, effluent treatment system, pressure and temperature control units, electrochemical test system, and safety-related components. The vessels are rated for a design pressure of 14.5 MPa (2100 psi) at 350°C (662°F). In this new facility, SCC initiation tests are planned on reverse U-bend and tubular specimens fabricated from multiple heats of Alloy 600MA, Alloy 600TT, and Alloy 690TT. Tubular specimens are to be statically loaded with periodic cycling at stress levels of 75%, 100%, and 150% of yield strength. Crack growth rate experiments are also planned on tubular fracture mechanics specimens under various stress intensities. The results from the autoclave tests will be utilized in model boiler experiments, and in the development of predictive models.

Mechanisms of Pb-induced SCC are also being studied. A variety of microscopic analysis techniques are being used to elucidate how Pb affects the protective films on the steam generator tubes, and to identify the particular valence states and compounds that do form. Preliminary tests were performed to investigate the effects of Pb on the electrochemical behavior of Alloy 600 in aqueous solutions. The results indicate that Pb ions would influence passivity of the Alloy 600 surface. Measurements of anodic polarization and potentiostatic electrochemical impedance spectroscopy were performed with deaerated solutions of pH 4.5 at 25 and 90°C, with or without addition of Pb (as 300 ppm Pb2+). The results indicated that lead was incorporated into the Alloy 600 specimen surface and enhanced electronic conductance. Incorporation of lead on the surface was examined for specimens tested at different electrochemical potentials (ECP). The specimens tested at cathodic potentials to the Pb/Pb2+ equilibrium showed Pb on the surface, while those tested at anodic potentials did not. The Pb/Pb2+ equilibrium ECP was calculated to be –0.210 and –0.259 V (vs. standard hydrogen electrode) at 25 and 90°C, respectively. Auger electron spectroscopy (AES) and x-ray photoelectron spectroscopy (XPS) were obtained on an Alloy 600 specimen surface that was treated potentiostatically in the above aqueous solutions. The AES data showed that depth concentration profiles of the major alloy elements (Ni, Cr, and Fe) are consistent with the bulk chemical analysis, except in the near surface region of 10-20 nm depth. Lead is observed up to 23 at.% in the surface layer of 20 nm for a specimen that was polarized at –0.27 V in the Pb-containing solution. XPS will provide information on oxidation states of the elements at the specimen surface within an effective depth of 3 nm. Additional AES and XPS will be performed on specimens treated under different electrochemical conditions. Further electrochemical tests, AES, and XPS will be applied to specimens produced at higher temperature and pressure in the autoclaves.

Steam generator tubes from McGuire Unit 1 and tube/tube sheet crevices were examined by destructive analysis. Intergranular cracks and attacks were observed at the secondary side, at the roll transition zone underneath crud deposits. To estimate the crevice chemistry that was present under operating conditions of the McGuire steam generator, we performed chemical and microstructural analyses for the deposits within the tube-to-tube sheet crevices in sections removed from the McGuire plant. A wide variety of elements were present in the crud deposits, including Fe, Ni, Cr, Al, Si, Mg, Cu, Ti, Mn, Ca, K, and S. The copper was present in the deposit as metallic copper. The presence of metallic Cu indicates that the electrochemical potential was below the Cu/Cu oxide equilibrium. The steam generator unit had operated initially with Ni-Cu moisture separator reheaters. It was not possible to identify whether lead was present. Additional microscopy is continuing to check for very local lead deposits at the crack and metal-metal oxide interfaces.

Corrosion and Mechanics of Materials: Light Water Reactors
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Last Modified: Thu, April 21, 2016 4:55 AM

 

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