Irradiation damage in 304 and 316 stainless steels: experimental investigation and modeling. Part I: Evolution of the microstructure

Pokor, C.; Bréchet, Y.; Dubuisson, Philippe; Massoud, J.P.; Barbu, A.

doi:10.1016/j.jnucmat.2003.11.007

Cited by 133 publications

(114 citation statements)

References 17 publications

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“…As a result, the material exhibits a substantial increase in yield stress (Fig. 2), reduction in ductility, swelling loss of corrosion resistance which may be damaging to the proper operation of the reactor [1][2][3][4][5][6][7]. For instance cracks were observed in bolts (Fig.…”

Section: State Of the Artmentioning

confidence: 99%

“…These two questions are currently under investigation up to the atomic scale on very short exposure time. HR(S)TEM (High Resolution (Scanning) Transmission Electron Microscopy) and ASTAR software clearly showed that the inner chromium rich oxide is a (FeCr) 3 O 4 spinel oxide growing with a cube/cube epitaxial crystallographic relationship with the base metal. HREELS (High Resolution Electron Energy Loss Spectroscopy) also clearly showed the possibility of quantifying the Fe 2+ /Fe 3+ ratio in the oxide and thus to get new informations on the passive property of the Cr rich film layer depending if this spinel is normal, intermediate or inversed ones.…”

mentioning

confidence: 99%

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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%

mentioning

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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“…Other recent studies also stated, without direct evidence, that black dots are believed to be interstitial type Frank loops in stainless steels irradiated with neutrons and ions. 12,19,20) The conclusion by Edwards et al was derived from experimental facts: coincidence in number density between the rel-rod images of Frank loops and the 2-beam dark field images of defect clusters, and comparisons of their data on the number density of Frank loops with literature data. A similar comparison is attempted in Fig.…”

Section: Dislocation Componentsmentioning

confidence: 99%

Evolution of Microstructure and Microchemistry in Cold-worked 316 Stainless Steels under PWR Irradiation

Fukuya¹,

Fujii²,

Nishioka³

et al. 2006

J. Nucl. Sci. Technol.

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The evolution of microstructures and microchemistry was examined by transmission electron microscopy in coldworked SUS 316 stainless steel components irradiated in a pressurized water reactor to 1-73 dpa at 565-596 K. Homogenous nucleation of dislocation loops, helium bubbles and 0 precipitates was detected. The dislocation loops consisted of a high density of Frank loops and black dots. The black dots are considered to be small Frank loops, some fraction of which could be vacancy-type. The size distribution showed a double peak, which remained unchanged up to 73 dpa, suggesting that the dislocation structure was saturated under a balance of nucleation and disappearance. The helium bubbles were extremely dense and fine, resulting in swelling of less than 0.1%. Such bubble structure was formed under irradiation with a high He/dpa ratio. The 0 precipitation was detected at doses higher than 4 dpa with an increasing density for higher doses. The measured radiation hardening was almost explained by these visible microstructural features. Radiation-induced segregation at grain boundaries was confirmed to continuously develop up to 73 dpa whereas no significant segregation could be detected at Frank loops.

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“…These materials constitute the internal structural components in Pressurized Water Reactors (PWR) and undergo neutron irradiation between 550 K and 600 K. The dose corresponding to the component lifetime (32 years of full time operation) is around 100 displacements per atom (dpa), while irradiation microstructures reach a steady state above doses of only 3 to 5 dpa [17][18][19]. The irradiation defects visible in TEM are interstitial Frank loops with a typical density between 0.2 and 1.0 10 23 m −3 and a mean diameter between 6 and 12 F o r P e e r R e v i e w O n l y 2 nm, depending on the material and the irradiation conditions.…”

mentioning

confidence: 99%

Atomic-scale plasticity in the presence of Frank loops

Nogaret

Robertson

Rodney

2007

Philosophical Magazine

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Irradiation damage in 304 and 316 stainless steels: experimental investigation and modeling. Part I: Evolution of the microstructure

Cited by 133 publications

References 17 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

Evolution of Microstructure and Microchemistry in Cold-worked 316 Stainless Steels under PWR Irradiation

Atomic-scale plasticity in the presence of Frank loops

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