Macroscopic and microscopic determinations of residual stresses in thin Oxide Dispersion Strengthened steel tubes

To improve the efficiency of components operating at high temperatures, many efforts are deployed to develop new materials. Oxide Dispersion Strengthened (ODS) materials could be used for heat exchangers or cladding tubes for the new GENIV nuclear reactors. This type of materials are composed with a metallic matrix (usually iron base alloy for nuclear applications or nickel base alloy for heat exchangers) reinforced by a distribution of nano-oxides. They are obtained by powder metallurgy and mechanical alloying. The creep resistance of these materials is excellent, and they usually exhibit a high tensile strength at room temperature. Depending on the cold working and/or the heat treatments, several types of microstructure can be obtained: recrystallised, stress relieved. One of the key challenges is to transform ODS materials into thin tubes (up to 500 microns thick) within a robust fabrication route while keeping the excellent mechanical properties. To prevent cracking during the process or to obtain a final product with low residual stresses, it is important to quantify the effect of the heat treatments on the release of internal stresses. The aim of this study is to show how residual stresses can be determined on different thin tubes using two complementary approaches: (i) macroscopic stresses determination in the tube using beam theory (small cuts along the longitudinal and circumferential directions and measurements of the deflection), (ii) stress determination from X-ray diffraction analyses (surface analyses, using "sin 2 Ψ" method with different hypothesis). Depending on the material and the heat treatment, residual stresses vary dramatically and can reach 800 MPa which is not far from the yield stress; comparisons between both methods are performed and suggestions are given in order to optimize the thermo-mechanical treatment of thin ODS tubes

Targeted Deletion of Kcne2 Causes Gastritis Cystica Profunda and Gastric Neoplasia

Gastric cancer is the second leading cause of cancer death worldwide. Predisposing factors include achlorhydria, Helicobacter pylori infection, oxyntic atrophy and TFF2-expressing metaplasia. In parietal cells, apical potassium channels comprising the KCNQ1 α subunit and the KCNE2 β subunit provide a K+ efflux current to facilitate gastric acid secretion by the apical H+K+ATPase. Accordingly, genetic deletion of murine Kcnq1 or Kcne2 impairs gastric acid secretion. Other evidence has suggested a role for KCNE2 in human gastric cancer cell proliferation, independent of its role in gastric acidification. Here, we demonstrate that 1-year-old Kcne2−/− mice in a pathogen-free environment all exhibit a severe gastric preneoplastic phenotype comprising gastritis cystica profunda, 6-fold increased stomach mass, increased Ki67 and nuclear Cyclin D1 expression, and TFF2- and cytokeratin 7-expressing metaplasia. Some Kcne2−/−mice also exhibited pyloric polypoid adenomas extending into the duodenum, and neoplastic invasion of thin walled vessels in the sub-mucosa. Finally, analysis of human gastric cancer tissue indicated reduced parietal cell KCNE2 expression. Together with previous findings, the results suggest KCNE2 disruption as a possible risk factor for gastric neoplasia

Simultaneous DAFS and XAFS analyses to evidence the Y- and Ti-species in nano-structured ODS steels

International audienceTo characterize the microstructure of neutron irradiated materials, various experimental tools are available at the present time in nuclear environment:SEM, TEM, APT, RBS and NRA… However, these techniques are local probes and bring information for an extremely limited sample volume which could be not representative of the whole volume. Therefore, a good statistic can only be obtained by performing many examinations which are very time consuming

Macroscopic and microscopic determinations of residual stresses in thin Oxide Dispersion Strengthened steel tubes

International audienceTo improve the efficiency of components operating at high temperatures, many efforts are deployed to develop new materials. Oxide Dispersion Strengthened (ODS) materials could be used for heat exchangers or cladding tubes for the new GENIV nuclear reactors. This type of materials are composed with a metallic matrix (usually iron base alloy for nuclear applications or nickel base alloy for heat exchangers) reinforced by a distribution of nano-oxides. They are obtained by powder metallurgy and mechanical alloying. The creep resistance of these materials is excellent, and they usually exhibit a high tensile strength at room temperature. Depending on the cold working and/or the heat treatments, several types of microstructure can be obtained: recrystallised, stress relieved. One of the key challenges is to transform ODS materials into thin tubes (up to 500 microns thick) within a robust fabrication route while keeping the excellent mechanical properties. To prevent cracking during the process or to obtain a final product with low residual stresses, it is important to quantify the effect of the heat treatments on the release of internal stresses. The aim of this study is to show how residual stresses can be determined on different thin tubes using two complementary approaches: (i) macroscopic stresses determination in the tube using beam theory (small cuts along the longitudinal and circumferential directions and measurements of the deflection), (ii) stress determination from X-ray diffraction analyses (surface analyses, using "sin 2 Ψ" method with different hypothesis). Depending on the material and the heat treatment, residual stresses vary dramatically and can reach 800 MPa which is not far from the yield stress; comparisons between both methods are performed and suggestions are given in order to optimize the thermo-mechanical treatment of thin ODS tubes

Characterizations with the MARS beamline (synchrotron SOLEIL) of materials irradiated in nuclear reactors

International audienceMARS (Multi-Analyses on Radioactive Samples) is the X-ray bending magnet beamline of the French synchrotron facility SOLEIL dedicated to the study of radioactive matter. The MARS beamline aims at extending the possibilities of synchrotron based X-ray characterizations towards a wider variety of radioactive elements (et61537;et61484;et61472;et61538;et61484;et61472;et61543; and n emitters). Thus, its specific and innovative infrastructure has been optimized to carry out analyses on radioactive materials with activities up to 18.5 GBq per sample. This beamline, which has been built thanks to a close partnership and support by the CEA, has been designed to provide X-rays in the energy range of 3.5 keV to 35 keV. Three main techniques are progressively proposed on MARS beamline transmission and high-resolution powder X-ray diffraction (respectively T-XRD and HR-XRD), X-ray absorption spectroscopy (XAS) and X-ray fluorescence (XRF).After the preliminary experiences performed on un-irradiated samples, this presentation deals with recent results obtained on the MARS beamline, thanks to very powerful and useful improvements brought to the experimental set-up of the beamline and to various materials irradiated in nuclear reactors Oxide dispersion-strengthened (ODS) steels at high doses and also Zr based alloys irradiated in Pressurized Water Reactors up to 7 PWR cycles.Results concerning secondary phases evolutions as a function of irradiation doses for both ODS and Zr based alloys will be presented using XRD, but also experiences using XAS especially on ODS will be given.Finally, future prospects and main objectives concerning the evolution of the beamline and studies on irradiated materials will be discussed

Understanding of Hybriding Mechanisms of Zircaloy-4 Alloy during Corrosion in PWR Simulated Conditions and Influence of Zirconium Hybrides on Zircaloy-4 Corrosion

Zirconium alloys are widely used as fuel claddings in Power Water Reactors. As they represent the first containment barrier to fission products, their mechanical integrity is essential for nuclear safety. During their corrosion in primary water, some of the hydrogen involved in the oxidation reaction with water ingresses into the alloy through the oxide layer. In the metallic matrix, once the solid solution limit is reached at the irradiation temperature, hydrogen precipitates as Zr hydrides mainly located just under the metal/oxide interface due to the thermal gradient across the cladding. As these hydrides may contribute to a larger oxide thickness and to a more fragile behaviour of the cladding, the minimization of hydrogen pick-up is required. Accordingly, since the Zircaloy-4 (Zr- 1.3Sn-0.2Fe-0.1Cr) alloy is known to be sensitive to this phenomenon, the understanding of its hydriding mechanism and of the influence of zirconium hydrides on its corrosion behaviour is needed. Regarding the study of the hydriding mechanism, isotopic exchanges were carried out in D2O environment at 360°C and led to the localization, in the oxide scales, of the limiting step for the hydrogen diffusion. To estimate an apparent diffusion coefficient of hydrogen in the oxide formed on Zircaloy-4, we firstly based on SIMS profiles and penetration depth of deuterium in the dense part of the oxide film. Secondly, ERDA estimation of the hydrogen content in zirconia and fusion measurement of the hydrogen content in both metal and oxide were used to estimate a hydrogen flux absorbed by the alloy and hence to deduce an apparent diffusion coefficient. Finally, these two methods lead to quite similar values (between 1.10-14 cm2/s and 6.10-14 cm²/s) which are in accordance with bibliography. Concerning the impact of hydrides on the corrosion of Zircaloy-4, several pre-hydrided and reference samples were corroded simultaneously in primary water at 360°C. The characterization of the pre-hydrided samples revealed some changes compared with the reference ones, as the presence of the Zr3O sub-oxide at the inner metal/oxide interface, a lower fraction of -ZrO2 in the oxide and a faster diffusion of oxygen species through grain boundaries of zirconia (TEM, μ-XRD, 18O isotopic experiments). Moreover, during oxidation, the hydrogen initially present in the hydride phase remains in the metallic matrix and leads to the allotropic transformation δ-ZrH1,66 ➔ ε-ZrH2

Investigation on the zirconia phase transition under irradiation

ACEZirconia, ZrO2, produced by the oxidation of zirconium alloys in nuclear reactors, possesses a high stability under neutron irradiation. No amorphisation of yttrium-stabilised zirconia has been observed even at high dpa values (≈100 dpa). In pure monoclinic zirconia, a phase transition monoclinic → cubic (tetragonal) induced by irradiation has already been observed. The aim of this work is to study in detail the mechanism responsible for this transition. For that purpose, different kinds of irradiations with electrons (to study point defects) and low energetic ions (to study clusters due to collision cascades) have been performed on zirconia samples. A local probe (Raman spectroscopy) and a non-local probe (grazing X-ray diffraction) have been used to characterise the phase formed during irradiation, which is clearly the tetragonal phase. For the ionic implantation, the grazing X-ray diffraction permits to separate effects due to the ballistic collisions and the implantation peak. Using this method, it was possible to show that the profile of the tetragonal phase was only linked to the dpa profile. This result associated to the results obtained by the Raman spectroscopy (broadening of Raman peaks) shows that the phase transition may be induced by clusters formed near the collision cascades