| 1. |
- Tejland, Pia, 1978, et al.
(författare)
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On the black oxide colour of zirconium alloys
- 2010
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115. ; 400:1, s. 79-83
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Tidskriftsartikel (refereegranskat)abstract
- Tubes of four zirconium alloys, used as cladding materials in light water nuclear reactors, were oxidized in an autoclave to produce oxide layers 1, 2 and 9 μm thick. The reflectance of the tubes was measured using a UV-vis-NIR spectrophotometer equipped with an integrating sphere detector. The 1 and 2 μm oxide tubes had a black appearance and a reflectance of only 10-12% in the visible region, whereas a dark grey appearing 9 μm oxide tube had 13-17% reflectance. The low reflectance was interpreted in terms of localized surface plasmon resonance in metallic particles embedded in the oxide layer, having a mean size of 40-70 nm. © 2010 Elsevier B.V. All rights reserved.
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| 2. |
- Eriksson, Johan, 1987, et al.
(författare)
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An atom probe tomography study of the chemistry of radiation-induced dislocation loops in Zircaloy-2 exposed to boiling water reactor operation
- 2021
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115. ; 550
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Tidskriftsartikel (refereegranskat)abstract
- This study is complementary to previous atom probe tomography (APT) studies of irradiation effects in the zirconium alloy Zircaloy-2. Using APT in voltage pulse mode, a difference in morphology was observed between clusters of Fe and Ni and clusters of Fe and Cr in Zircaloy-2 exposed to a high fast neutron fluence in a commercial boiling water reactor. The Fe–Ni clusters were disc-shaped with a diameter of 5–15 nm, whereas the Fe–Cr clusters were spheroidal with a diameter of approximately 5 nm. Both types of clusters appeared to be located at irradiation-induced -type dislocation loops aligned in layers normal to the -direction. The concentration of Fe was higher in the Fe–Cr clusters than in the Fe–Ni clusters. The dilute Fe–Ni clusters, which seem to be segregation of Fe and Ni inside the loops, had formed on all three families of first-order prismatic planes with some deviation from perfect -axis alignment. The Fe–Cr clusters might be very small precipitates with a nucleation associated with the loops.
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| 3. |
- Eriksson, Johan, 1987, et al.
(författare)
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Nanoscale chemistry of Zircaloy-2 exposed to three and nine annual cycles of boiling water reactor operation — an atom probe tomography study
- 2022
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115. ; 561
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Tidskriftsartikel (refereegranskat)abstract
- Atom probe tomography was used in this work to study the metal close to the metal/oxide interface in the zirconium alloy Zircaloy-2 exposed to three and nine annual cycles of operation in a commercial boiling water reactor. The two exposure times correspond to before and after the onset of acceleration in corrosion, hydrogen pickup, and growth. The alloying elements Sn, Fe, Cr, and Ni were observed to be redistributed after exposure. After both three and nine cycles, clusters containing Fe and Cr and typically of a spheroidal shape with an approximate diameter of 5 nm were observed to be located in layers presumed to be layers of -loops. On average, the cluster number density was slightly higher after nine cycles, with larger and more Cr-rich clusters. However, there were large grain-to-grain variations, which were larger than the differences between the two exposure times. Ni was only occasionally observed in the clusters. Sn was observed to be slightly enriched in the Fe–Cr clusters, but the Sn concentration was higher between than inside the layers of clusters. After nine cycles, clusters of Sn were detected in regions that were depleted of Fe and Cr. Enrichment of Sn, Fe, and Ni at features that appeared to be -component loops was observed after nine cycles, whereas no such features were observed after three cycles. Enrichment of Sn and Fe, and small amounts of Cr and Ni, was observed at grain boundaries after both exposure times. After three cycles, a partially dissolved second phase particle of Zr(Fe,Cr)2 type that contained about ten times more Cr than Fe was observed.
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| 4. |
- Sundell, Gustav, 1985, et al.
(författare)
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Redistribution of alloying elements in Zircaloy-2 after in-reactor exposure
- 2014
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115. ; 454:1-3, s. 178-185
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Tidskriftsartikel (refereegranskat)abstract
- An atom probe tomography study of the microstructure of a Zircaloy-2 material subjected to 9 annual cycles of BWR exposure has been conducted. Upon dissolution of secondary phase particles, Fe and Cr are seen to reprecipitate in large numbers of clusters and particles of 1-5 nm sizes throughout the Zr metal matrix. Fe and Sn were observed to segregate to ring-shaped features in the metal that are interpreted to be -component vacancy loops. This implies that these two elements play a major role in the irradiation growth phenomenon in Zr alloys, which is believed to be caused by the formation of -loops. Similarly to autoclave-corroded Zr alloys, the formation of a sub-oxide layer of approximate composition ZrO was observed. On the other hand, no oxygen saturated metal phase was detected underneath the oxide scale.
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| 5. |
- Tejland, Pia, 1978, et al.
(författare)
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Detailed analysis of the microstructure of the metal/oxide interface region in Zircaloy-2 after autoclave corrosion testing
- 2011
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Ingår i: ASTM Special Technical Publication. 16th International Symposium on Zirconium in the Nuclear Industry, Chengdu, Sinchuan Province, 9-13 May 2010. - 0066-0558. - 9780803175150 ; 1529 STP, s. 595-617
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Konferensbidrag (refereegranskat)abstract
- Two varieties of Zircaloy-2, with different second phase particle (SPP) size distributions and different corrosion resistance, were oxidized in a steam autoclave. Transmission electron microscopy (TEM) of large thin-foil cross-sections of the oxide and the adjacent metal shows an undulating metal/oxide interface in both materials with a periodicity of slightly less than 1 μm and an amplitude of around 100 nm. The SPPs oxidize slower than the surrounding metal, and the absence of volume increase leads to void and crack formation as the SPPs become embedded in the oxide. On SPP oxidation, iron diffuses out of the particles into the surrounding oxide. A sub-oxide with an oxygen content of approximately 50 at. % and a layer thickness of about 200 nm was observed close to the metal/oxide interface. There is a 200 nm oxygen concentration gradient into the metal, from the level close to the sub-oxide of about 30 at. % down to a few atomic percent. All tin in the matrix is incorporated in the sub-oxide, and no segregation to the metal/oxide interface was found.
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| 6. |
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| 7. |
- Tejland, Pia, 1978
(författare)
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Microstructure Investigation of the Oxidation Process in Zircaloy-2 - The Effect of Intermetallic Particle Size
- 2012
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Zirconium alloys are widely used in nuclear reactors as fuel cladding tubes because of their low thermal neutron capture cross-section, good corrosion properties and satisfactory mechanical properties. With an improved corrosion resistance the alloys could be used for much longer times in the reactors, increasing fuel burn-up and decreasing the amount of radioactive waste. Therefore it is of great importance to try to understand the mechanisms of the oxidation process in these alloys.In this study the oxidation behavior in steam autoclave of Zircaloy-2, an alloy used primarily in boiling water reactors, is studied. Special emphasis is put on the role of the intermetallic second phase particles (SPPs) containing iron, chromium and nickel and with a typical size of 50 nm. The main method for investigations has been transmission electron microscopy in combination with energy dispersive X-ray spectroscopy. Also atom probe tomography and spectrophotometry have been used. Focus has been on the microstructure of the oxide and the metal/oxide interface zone. It was found that the SPPs oxidize slower than the surrounding metal, and that the absent volume increase leads to void and crack formation as the SPPs become embedded in the oxide. On SPP oxidation, iron diffuses out of the particle into the surrounding oxide.The metal/oxide interface was found to undulate on a micrometer scale. This undulation gives rise to large stresses perpendicular to the metal/oxide interface. In the oxide, above wave crests, lateral cracks are formed, and it is shown that un-oxidized SPPs embedded in the oxide act as nucleation sites for these cracks. Therefore, a material with many small SPPs has more lateral cracks than a material with few large SPPs.Adjacent to the oxide often a sub-oxide layer (~ZrO) is found, with varying thickness also in the same specimen. One sub-oxide layer with an average oxygen content of ~55 at. % was found to consist of fingers with ~60 at. % oxygen with a diameter of 5 nm and a length of 50 nm, penetrating into regions of ~50 at. % oxygen. From the oxide, oxygen diffuses into the metal and it was found that the width of the oxygen diffusion profile varies, with wider profiles underneath delayed parts of the interface. An oxygen enriched phase with ~30 at. % oxygen was found in some of the specimens. Evidence of extensive plastic deformation in the metal underneath the oxide scale was found in the form of twinning, dislocation tangles and patches, cell formation and sub-grain formation. The heavily deformed layer is a few µm thick and no obvious difference could be seen underneath different oxide thicknesses or between alloys with different strength. The black oxide color of zirconium alloys has been studied using spectrophotometry. The conclusion is that the reason for the black appearance of oxidized zirconium alloys is the excitation of localized surface plasmon resonances in the metallic SPPs embedded in the oxide layer.
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| 8. |
- Tejland, Pia, 1978
(författare)
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Microstructure of Zirconium Alloys Oxidized in Steam
- 2009
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Licentiatavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Zirconium alloys are widely used in nuclear reactors as fuel cladding tubes because of their low thermal neutron capture cross-section, good corrosion properties and satisfactory mechanical properties. However, the main limiting factor of these materials is the oxidation behavior. With an improved corrosion resistance the alloys could be used for much longer times in the reactors. Therefore it is of great importance to try tounderstand the mechanisms of the oxidation process in order to improve the alloys for a better performance.In this study the oxidation behavior of zirconium alloys is studied, and special emphasis is put on the role of the intermetallic second phase particles (SPPs) containing iron,chromium and nickel and with a typical size of 50 nm. The alloys chosen for this study are materials that have long been used in reactors and for which long-term in-reactordata exists. The main method for investigations has been transmission electron microscopy (TEM) in combination with energy dispersive X-ray spectroscopy.As a first step, methods were developed for manufacturing large thin foil TEM specimens, containing the metal/oxide interface, using the focused ion beam in-situ liftouttechnique, and also for imaging the SPPs. The particles in the oxide are hard to spot with conventional bright field TEM due to substantial crystallographic contrast from thesmall oxide grain size. It was concluded that the best way to image the particles in the oxide is by using a high angle annular dark field detector in TEM.The morphology of the oxide and the metal/oxide interface was studied. It was found that the interface is undulating on a micrometer scale and sub-oxide was found close to the interface.The oxidation of SPPs was studied and it was concluded that the SPPs oxidize slower than the surrounding metal, and that the absent volume increase leads to void and crackformation as the SPPs become embedded in the oxide. On SPP oxidation, iron diffuses out of the particle into the surrounding oxide.The black oxide color of zirconium alloys has also been studied, using spectrophotometry. The conclusion is that the reason for the black appearance of oxidized zirconium alloys is the excitation of localized surface plasmon resonances in the metallic SPPs embedded in the oxide layer. When the oxide grows thicker it turns greyish as SPPs in the outer part of the oxide oxidize so that light has to travel through a thicker, partly cracked oxide before meeting absorbing particles.
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| 9. |
- Tejland, Pia, 1978, et al.
(författare)
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Origin and effect of lateral cracks in oxide scales formed on zirconium alloys
- 2012
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115. ; 430:1-3, s. 64-71
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Tidskriftsartikel (refereegranskat)abstract
- Two varieties of Zircaloy-2, with different second phase particle (SPP) size distributions and different corrosion resistance, were oxidized in a steam autoclave. Transmission electron microscopy was used for investigation of the fine-scale lateral cracks present in the oxide scales. Crack quantification was performed and the number of cracks was correlated with the number of SPPs. A mechanism for crack formation is presented, in which the driving force is the local tensile stresses in the oxide close to the oxide/metal interface, and the initiation sites are un-oxidized SPPs located within this stress field.
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| 10. |
- Tejland, Pia, 1978, et al.
(författare)
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Oxidation induced localized creep deformation in Zircaloy-2
- 2014
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115. ; 444:1-3, s. 30-34
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Tidskriftsartikel (refereegranskat)abstract
- Extensive plastic deformation in the metal underneath the oxide scale in autoclave tested Zircaloy-2 was studied using transmission electron microscopy (TEM). It was concluded that the plastic deformation is created by creep during oxidation, and is not caused by surface treatment, sample preparation or cooling from autoclave temperatures. Evidence of large strains was found in the form of dislocation tangles, dis- location patches and sub-grain formation, and also indications of twinning were found. The heavily deformed layer is around a few lm thick and no obvious difference could be seen between alloys with different strength or different oxide thickness.
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| 11. |
- Tejland, Pia, 1978, et al.
(författare)
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Oxidation Mechanism in Zircaloy- 2—The Effect of SPP Size Distribution
- 2015
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Ingår i: ASTM Special Technical Publication. - 0066-0558. - 9780803175297 ; 1543, s. 373-403
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Konferensbidrag (refereegranskat)abstract
- The metal/oxide interface region in Zircaloy-2 oxidized in autoclave was studied with transmission electron microscopy (TEM) and atom probe tomography. In addition to waviness on the micrometer scale the metal/oxide interface was found to have irregularities on a finer scale, and metal islands were found especially at metal hills (delayed parts of the oxidation front). The thickness of the sub-oxide layer varies considerably along the interface in the same sample, from 100 to virtually 0 nm. The sub-oxide composition may vary on a very fine scale (down to 5nm), and it can sometimes be a mixture of sub-oxides with different oxygen content. The metal matrix in contact with the sub-oxide is saturated with up to 32 at. % oxygen, and the oxygen diffusion profile in the metal is in approximate agreement with literature data for pure Zr. However, the diffusion length appears to be somewhat larger at interface metal hills than under valleys, probably for both geometrical and stress state reasons. Hydride precipitates, hardly visible in conventional TEM, give a good image contrast when employing high angle annular dark field imaging. A model for the oxidation process is presented, where the creep deformation of the metal close to the interface and the formation of lateral cracks in the oxide are of highest importance. The effect of second phase particle (SPP) size is suggested to be twofold: Small and numerous SPPs give a stronger metal and therefore higher stress in the oxide. Small SPPs also nucleate many more lateral cracks in the oxide, which gives a weaker oxide. Together this leads to formation of large cracks associated with transition in the oxidation rate at an earlier time than for a material with larger and fewer SPPs, and thereby a higher oxidation rate.
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| 12. |
- Topping, Matthew, et al.
(författare)
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The effect of iron on dislocation evolution in model and commercial zirconium alloys
- 2018
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Ingår i: ASTM Special Technical Publication. - 0066-0558. ; STP 1597, s. 796-822
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Konferensbidrag (refereegranskat)abstract
- Although the evolution of irradiation-induced dislocation loops has been well correlated with irradiation-induced growth phenomena, the effect of alloying elements on this evolution remains elusive, especially at low fluences. To develop a more mechanistic understanding of the role iron has on loop formation, we used state-of-the-art techniques to study a proton-irradiated Zr-0.1Fe alloy and proton- and neutron-irradiated Zircaloy-2. The two alloys were irradiated with 2-MeV protons up to 7 dpa at 350°C and Zircaloy-2 up to 14.7 × 1025n • m-2, approximately 24 dpa, in a boiling water reactor at approximately 300°C. Baseline transmission electron microscopy showed that the Zr3Fe secondary-phase particles in the binary system were larger and fewer in number than the Zr (Fe, Cr)2and Zr2(Fe, Ni) particles in Zircaloy-2. An analysis of the irradiated binary alloy revealed only limited dissolution of Ze3Fe, suggesting little dispersion of iron into the matrix, while at the same time a higher 〈a〉-loop density was observed compared with Zircaloy-2 at equivalent proton dose levels. We also found that the redistribution of iron during irradiation led to the formation of iron nanoclusters. A delay in the onset of 〈c〉-loop nucleation in proton-irradiated Zircaloy-2 compared with the binary alloy was observed. The effect of iron redistributed from secondary-phase particles because of dissolution on the density and morphology of 〈a〉 and 〈c〉 loops is described. The implication this may have on irradiation-induced growth of zirconium fuel cladding is also discussed.
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