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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. Source


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. Source


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. Source


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. Source


Wang J.,CAS Shenyang Institute of Metal Research | Li X.,CAS Shenyang Institute of Metal Research | Li X.,State Nuclear Power Plant Service Company | 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. Source

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