Staehle Consulting

North Oaks, MN, United States

Staehle Consulting

North Oaks, MN, United States
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Wang J.,CAS Shenyang Institute of Metal Research | Li X.,CAS Shenyang Institute of Metal Research | Li X.,State Nuclear Power Plant Service Co. | Huang F.,CAS Shenyang Institute of Metal Research | And 2 more authors.
Corrosion | Year: 2014

Corrosion behaviors of UNS N06690TT and N08800SN in simulated pressurized water reactor (PWR) primary water containing three concentrations of dissolved oxygen (DO) was studied by open-circuit potential (OCP), electrochemical impedance spectroscopy (EIS), scanning electron microscopy (SEM), x-ray photoelectron spectroscopy (XPS), and transmission electron microscopy (TEM). These results have shown that DO in simulated PWR primary water produces a strong effect on Cr dissolution. At DO < 0.01 ppm, the electrochemical impedance of N06690TT at low frequency gradually increases and becomes larger than that of N08800SN, and N06690TT shows slightly better corrosion resistance than N08800SN. At DO ≥ 0.1 ppm, the electrochemical impedance of N06690TT at low frequency rapidly decreases and becomes lower than that of N08800SN with the increase of immersion time; N08800SN shows much better corrosion resistance than N06690TT. The spinel oxides NiFe2O4 or Fe 3O4 can form in the inner oxide layer on N08800SN due to relatively balanced ratio of Fe, Ni, which leads to the relatively stable corrosion resistance. Therefore, when DO in PWR primary water is 0.1 ppm or more, it is more suitable to choose N08800SN rather than N06690TT as the steam generator (SG) tubing materials. © 2014, NACE International.


Staehle R.W.,Staehle Consulting
Corrosion Reviews | Year: 2010

The purpose of this discussion is to assess the implications of fifteen years of studies of stress corrosion cracking (SCC) using the Analytical Electron Transmission Microscope (ATEM). This work has considered mainly Fe-Cr-Ni alloys of types used in water cooled nuclear power plants in the range of 250-350°C with hydrogenated and oxygenated water and with some contaminated environments. The objectives of this work have been mainly to determine causes of failures usually with respect to impurities in the water, faulty materials, geometric effects, irradiation, and from other sources. While the work, nominally, has not been directed toward mechanistic studies, the workers in this field have conducted some very good science enabling important insights into critical processes that affect predicting the course of SCC. Among the most important of the scientific advances has been showing that tips of cracks are in the range of 1-5 nm wide and are not in the range of large crack tip opening displacements (CTODs) that are calculated to be in the range of 2500-5000 nm wide. This finding of such narrow crack tips represents a "paradigm shift" in considering mechanisms of SCC. These very thin or "tight cracks", are so narrow that they must be treated on a molecular basis and not by a continuum basis. In view of this geometry, these crack tips are now referred to as "tight cracks" or as "molecular cracks." Second, these ATEM studies have shown that metal at the crack tip is often enriched in the noble component, Ni. This finding seems more likely in aborted cracks and less likely in propagating cracks although there is disagreement among researchers. Ni enrichment is not so common in high nickel alloys. This finding is a major advance and raises important questions about the resistance of all alloys now being used from the Fe-Cr-Ni alloy system. Also, this finding provides an avenue for conducting useful mechanistic research directed toward predicting long lifetimes. There is evidence suggesting that the advance of SCC is a brittle process and is not associated with breaking of passive films. Studies of the mature SCC after cracking at the crack tip, show that the oxides formed do not come from the crack tip but rather from in situ oxidation. From an engineering point of view the ATEM work has provided important answers to effects of contaminants and water chemistry. Recommendations for future scientific work with the ATEM include a thorough definition of dislocation arrays at the crack tip, detailed study of the metal composition ahead of the crack tip, and detailed characterization of the physical and chemical features of the tips of the tight cracks.


Staehle R.W.,Staehle Consulting
15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors 2011 | Year: 2011

The purpose of this discussion is to describe the Quantitative Micro-Nano (QMN) project. The QMN approach is directed toward predicting SCC in components of the primary and secondary systems of water-cooled nuclear plants. Two week-long meetings, QMN-1 and QMN-2 have been held in the last two years, 2010 and 2011. A third meeting is being planned for 2012 in June. The QMN program is based on the five segments of initiation and propagation: initial condition, precursors, incubation, proto-cracks, and propagation of larger cracks. Each of these segments is the subject of specific meetings where the detailed atomic processes are described and quantified. These mechanistic elements will then be synthesized into overall models. This same approach is being taken in structural biology and electronic materials. QMN benefits from presently available methods for characterizing structure and chemistry at an atomic level and then manipulating the results with arrays of multi-particle computer models. Participation in the QMN meetings is organized by a scientific advisory committee that recommends topics and presenters.


Staehle R.W.,Staehle Consulting
Corrosion Reviews | Year: 2010

The subject of proactivity with respect to assuring the future safety, reliability, and economical operation of water-cooled nuclear plants is discussed. This subject is particularly important as these plants are expected to operate for possibly twice their present lives, and there is no experience with failures that might occur in the future for such reactor lifetimes. Proactive planning is the opposite of the reactive response in the past where failures were investigated after they had occurred and not predicted, as they could have been, before they occurred. Proactive research should be aimed at preventing or minimizing especially serious failures in materials and components before such failures occur. This present analysis of the "anatomy of proactivity" considers proactivity more comprehensively than early previous work on the subject. Predicting future failures needs to include considerations of failure modes, influence of management in causing failures, need for improved NDE to identify the initiation of future indications of failures, actions by individuals, and the importance of the organizations in nuclear power. This broader consideration is "comprehensive proactivity." It is the anatomy of this "comprehensive proactivity" that is discussed here.


Ru X.,Harbin Engineering University | Ru X.,Nuclear Power Institute of China | Staehle R.W.,Staehle Consulting
Corrosion | Year: 2013

This review assesses past experiences from superheated fossil plants, supercritical fossil plants, superheated nuclear plants, and light water reactors from the late 1940s until the present. Data from the development and operation of these plants are directly applicable to supercritical water reactor (SCWR) plants being developed currently. This past work can be applied to the development of current designs in the choice of materials, temperature dependencies, effects of stress, and effects of environments on materials. Some of the past data from light water reactor (LWR) technology can be extrapolated into the present SCWR regimes. The past data are in good agreement among the various previous investigators. These past data are considered with respect to specific components in SCWR: fuel cladding, reactor structurals, reactor vessels, and feed water heaters. Choosing materials for the SCWR applications must recognize that the materials at nominal outlet temperatures are in a dynamic thermal range, i.e., in the nuclear superheat and fossil superheat range, the atomic structures of materials change significantly and change properties such as ductility. It is also possible that compositions and structures of grain boundaries can change, for example, the susceptibility to SCC. The surface temperatures on fuel cladding will be significantly higher than the outlet temperature, and both temperatures will exceed, substantially, the outlet temperatures of present water-cooled plants. Past isothermal data directed toward core structures may not be relevant to the same alloy as fuel cladding. The outlet temperature is useful for considering core structural materials but not for fuel elements, owing to an inevitably high film drop. SCW environments are expected to produce extensive SCC, which differs from past experience. Finally, past data were reanalyzed and additional useful insights were obtained. © 2013, NACE International.


Ru X.,Harbin Engineering University | Ru X.,Nuclear Power Institute of China | Staehle R.W.,Staehle Consulting
Corrosion | Year: 2013

This review assesses past experiences from superheated fossil plants, supercritical fossil plants, superheated nuclear plants, and light water reactors from the late 1940s until the present. Data from the development and operation of these plants are directly applicable to supercritical water reactor (SCWR) plants being developed currently. This past work can be applied to the development of current designs in the choice of materials, temperature dependencies, effects of stress, and effects of environments on materials. Some of the past data from light water reactor (LWR) technology can be extrapolated into the present SCWR regimes. The past data are in good agreement among the various previous investigators. These past data are considered with respect to specific components in SCWR: fuel cladding, reactor structurals, reactor vessels, and feedwater heaters. Choosing materials for the SCWR applications must recognize that the materials at nominal outlet temperatures are in a dynamic thermal range, i.e., in the nuclear superheat and fossil superheat range, the atomic structures of materials change significantly and change properties such as ductility. It is also possible that compositions and structures of grain boundaries can change, for example, the susceptibility to SCC. The surface temperatures on fuel cladding will be significantly higher than the outlet temperature, and both temperatures will exceed, substantially, the outlet temperatures of present water-cooled plants. Past isothermal data directed toward core structures may not be relevant to the same alloy as fuel cladding. The outlet temperature is useful for considering core structural materials but not for fuel elements, owing to an inevitably high film drop. SCW environments are expected to produce extensive SCC, which differs from past experience. Finally, past data were reanalyzed and additional useful insights were obtained. © 2013, NACE International.


Arioka K.,Institute of Nuclear Safety Systems Inc. | Staehle R.W.,Staehle Consulting | Yamada T.,Institute of Nuclear Safety Systems Inc. | Miyamoto T.,Kobe Material Testing Laboratory Co. | Terachi T.,Institute of Nuclear Safety Systems Inc.
Corrosion | Year: 2016

The purpose of this work is to understand quantitative processes which underlie the initiation of stress corrosion cracking (SCC) of cold-worked, thermally treated Alloy 690 after longterm exposures in high-temperature water. Long-term stress corrosion cracking initiation tests up to 34,484 h have been performed on cold-worked, thermally treated Alloy 690 under static load conditions using 0.5T compact tension specimens with a shallow depth of the precrack (0.1 mm to approximately 0.3 mm) and blunt notch compact tension type specimens. These specimens were exposed in the primary coolant environments of pressurized water reactors under static load conditions both at 320°C and 360°C. Three important patterns were observed. First, intergranular cracking observed from the shallow precracks after 21,838 h at 320°C and 20,653 h at 360°C, respectively, in 20% cold-worked Alloy TT690 in primary water. Second, clear evidence of cavities were identified ahead of the SCC-tip after the tests both at 360°C and 320°C. Cavities seem to result from condensations of vacancies induced by cold work, which were driven by stress gradients judging from the distributions of the population of cavities. Third, oxidation occurred inside the cavities near the SCC-tips before the advance of SCC. Bonding strength at grain boundaries is assumed to weaken as a result of the formation of cavities and oxidation inside cavities during the incubation of long-term SCC. A model is proposed for the initiation of SCC after longterm operation with cold-worked, thermally treated Alloy 690 in high-temperature water, involving the combination of local corrosion and the formation of cavities. These result from the collapses of vacancies, which seem to arise from the initiation of SCC after long times in high-temperature water. © 2016, NACE International.

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