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

INVESTIGATION OF NEUTRON RADIATION EFFECTS ON THE MECHANICAL BEHAVIOR OF RECRYSTALLIZED ZIRCONIUM ALLOYS

International audienceNeutron radiation induces important changes in the mechanical behavior of recrystallized zirconium alloys used as fuel cladding tube. The neutron radiation effects on the mechanical behavior for internal pressure test performed at 350DC have been investigated using a specific analysis in terms of isotropic hardening, kinematic hardening and viscous stress. The impact of irradiation has been interpreted in terms of microscopic deformation mechanisms observed by Transmission Electron Microscopy (TEM). It is proposed that because of the localization of the plastic deformation inside channels and because of the only activation of basal channeling, the kinematic hardening must be strong in irradiated zirconium alloys. A simple unified phenomenological modeling is also used in order to have a coherent description of the radiation effects on the mechanical behavior

INVESTIGATION OF NEUTRON RADIATION EFFECTS ON THE MECHANICAL BEHAVIOR OF RECRYSTALLIZED ZIRCONIUM ALLOYS

International audienceNeutron radiation induces important changes in the mechanical behavior of recrystallized zirconium alloys used as fuel cladding tube. The neutron radiation effects on the mechanical behavior for internal pressure test performed at 350DC have been investigated using a specific analysis in terms of isotropic hardening, kinematic hardening and viscous stress. The impact of irradiation has been interpreted in terms of microscopic deformation mechanisms observed by Transmission Electron Microscopy (TEM). It is proposed that because of the localization of the plastic deformation inside channels and because of the only activation of basal channeling, the kinematic hardening must be strong in irradiated zirconium alloys. A simple unified phenomenological modeling is also used in order to have a coherent description of the radiation effects on the mechanical behavior

Investigation of Transformation Temperatures, Microstructure and Shape Memory Properties of NiTi, NiTiZr and NiTiHf Alloys

The influence of zirconium and hafnium additions in substitution to titanium, is investigated on TiNi shape memory alloy. Microstructural and thermomechanical studies are conducted on an arc-melted Ti38Ni50Hf12 alloy. Results are compared with those of an equiatomic TiNi. We confirm the increase of the phase transformation temperatures due to the additions of Zr and Hf. The presence of oxides in the cast alloy was detected. The lattice parameter is measured from a residue of electroextraction. It was found that a0=1,1547±0,0001nm. A stoichiometry of (Ti, Hf) 4Ni2 Ox (with x≤1) is suggested. It appears that this ternary alloy exhibit an improved shape recovery when it is pre-strained in austenite state (for Ms < T < As) . In this case it displays 100% of recovery for 4% of deformation

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