The composition and morphology of oxide films formed on type 304 stainless steel in lithiated high temperature water

Tapping, R.L.; Davidson, R. D.; McAlpine, E.; Lister, D. H.

doi:10.1016/0010-938x(86)90024-7

Cited by 92 publications

(34 citation statements)

References 12 publications

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“…All the open literature agreed that oxide layer formed in PWR primary environment has duplex structure while its chemistry and crystal structure remain controversial [87][88][89][90][91][92][93][94][95][96][97][98][99][100][101]. Even if it is well established that the outer layer is composed of magnetite Fe 3 O 4 , the inner layer composition is reported to be either a fcc phase enriched in chromium [76,77,[90][91][92][93][94][95][96][97][98][99][100][101], or an hexagonal phase (Cr 2 O 3 ) [87,92,93], or even a mixture of both phases [92]. What's more the early stages of oxidation process (1 min to 5 hours) without applying stress was still not investigated since the already published results describe oxides formed after long exposure times (up to 1,000 hours).…”

Section: State Of the Artmentioning

confidence: 99%

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Legras¹,

Volgin²,

Radiguet³

et al. 2017

JMSE-B

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Abstract:Even if temperature, pressure and chemistry of the cooling water are not very high and aggressive, materials used in PWRs (Pressurized Water Reactors) are exposed to different degradation mechanisms. One of the main goals of the research programs in this field is to develop physical model of the mechanisms down to the atomic scale. Such approach needs a clear description and understanding of the degradation mechanisms at the same small scale. This paper illustrates the benefits of different microscopies and of their last improvements up to the promising possibilities of monochromated and aberrations corrected TEM/STEM. A specific focus is placed on four different degradation mechanisms observed in austenitic stainless steel: irradiation ageing, corrosion fatigue, stress corrosion cracking and corrosion.

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Section: State Of the Artmentioning

confidence: 99%

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Legras¹,

Volgin²,

Radiguet³

et al. 2017

JMSE-B

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“…However, both of X-ray and electron diffraction measurements did not show this inner chromium rich oxide layer, then it is thought that chromium rich oxide is amorphous or consists of too small crystals to obtain the diffraction rays. According to Tapping et al,15) the inner layer of oxide film formed on type 304 stainless steel is chromium rich amorphous. Although the materials are different, the present result is consistent with that on stainless steel, which consists of iron, chromium and nickel as same as Alloy 600.…”

Section: Relation Between Hydrogen Content and Oxide Film Structurementioning

confidence: 99%

Influence of Dissolved Hydrogen on Structure of Oxide Film on Alloy 600 Formed in Primary Water of Pressurized Water Reactors

Terachi¹,

Totsuka²,

Yamada³

et al. 2003

J. Nucl. Sci. Technol.

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In order to investigate the relationship between the susceptibility of primary water stress corrosion cracking (PWSCC) in Alloy 600 and the content of dissolved hydrogen (DH) in the primary water of pressurized water reactors (PWR), structural analysis of oxide films formed under four different DH conditions in simulated primary water of PWR was carried out using a grazing incidence X-ray diffractometer (GIXRD), a scanning electron microscope (SEM) and a transmission electron microscope (TEM). In particular, to perform accurate analysis of the thin oxide films, the synchrotron radiation of SPring-8 was used for GIXRD.It has been observed that the oxide film is mainly composed of nickel oxide, under the condition without hydrogen. On the other hand, needle-like oxides are formed at 1.0 ppm of DH. In the environment of 2.75 ppm of DH, the oxide film has thin spinel structures. From these results and phase diagram consideration, the condition around 1.0 ppm of DH corresponds to the boundary between stable NiO and spinel oxides, and also to the peak range of PWSCC susceptibility. This suggests that the boundary between NiO and spinel oxides may affect the SCC susceptibility.

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“…[52][53][54][55][56][57][58] The inner layer consists of chromium-rich spinel Ni x Cr y Fe 3-x-y O 4 with nonstoichiometric composition; the actual composition of spinels varies with environmental conditions. Da Cunha Belo et al 56 O 4 , in the inner layer formed on Types 304 and 316 SS exposed at 288°C under conditions of NWC or hydrogen water chemistry (HWC).…”

Section: Surface Oxide Filmmentioning

confidence: 99%

Mechanism and estimation of fatigue crack initiation in austenitic stainless steels in LWR environments.

Chopra¹

2002

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The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life (e-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of fatigue crack initiation in austenitic stainless steels in LWR coolant environments. The existing fatigue e-N data have been evaluated to establish the effects of key material, loading, and environmental parameters (such as steel type, strain range, strain rate, temperature, dissolved-oxygen level in water, and flow rate) on the fatigue lives of these steels. Statistical models are presented for estimating the fatigue e-N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. The influence of reactor environments on the mechanism of fatigue crack initiation in these steels is also discussed. Executive SummarySection III, Subsection NB, of the ASME Boiler and Pressure Vessel Code contains rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III specify the Code design fatigue curves for applicable structural materials. However, Section III, Subsection NB-3121, of the Code states that effects of the coolant environment on fatigue resistance of a material were not intended to be addressed in these design curves. Therefore, the effects of environment on fatigue resistance of materials used in operating pressurized water reactor (PWR) and boiling water reactor (BWR) plants, whose primary-coolant pressure boundary components were designed in accordance with the Code, are uncertain.The current Section-III design fatigue curves of the ASME Code were based primarily on strain-controlled fatigue tests of small polished specimens at room temperature in air. Best-fit curves to the experimental test data were first adjusted to account for the effects of mean stress and then lowered by a factor of 2 on stress and 20 on cycles (whichever was more conservative) to obtain the design fatigue curves. These factors are not safety margins but rather adjustment factors that must be applied to experimental data to obtain estimates of the lives of components. They were not intended to address the effects of the coolant environment on fatigue life. Recent fatigue-strain-vs.-life (e-N) data obtained in the U.S. and Japan demonstrate that light water reactor (LWR) environments can have potentially significant effects on the fatigue resistance of materials. Specimen lives obtained from tests in simulated LWR environments can be much shorte...

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The composition and morphology of oxide films formed on type 304 stainless steel in lithiated high temperature water

Cited by 92 publications

References 12 publications

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Influence of Dissolved Hydrogen on Structure of Oxide Film on Alloy 600 Formed in Primary Water of Pressurized Water Reactors

Mechanism and estimation of fatigue crack initiation in austenitic stainless steels in LWR environments.

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