Download or read book Microstructural Characterization of Oxides Formed on Model Zr Alloys Using Synchrotron Radiation written by A. T. Motta and published by . This book was released on 2008 with total page 21 pages. Available in PDF, EPUB and Kindle. Book excerpt: To understand how alloy chemistry and microstructure impact corrosion performance, oxide layers formed at different stages of corrosion on various model zirconium alloys (Zr-xFe-yCr, Zr-xCu-yMo, for various x, y) and control materials (pure Zr, Zircaloy-4) were examined to determine their structure and the connection of such structure to corrosion kinetics and oxide stability. Microbeam synchrotron radiation diffraction and fluorescence of oxide cross sections were used to determine the oxide phases present, grain size, and orientation relationships as a function of distance from the oxide-metal interface. The results show a wide variation of corrosion behavior among the alloys, in terms of the pretransition corrosion kinetics and in terms of the oxide susceptibility to breakaway corrosion. The alloys that exhibited protective behavior at 500°C also were protective during 360°C corrosion testing. The Zr-0.4Fe-0.2Cr model ternary alloy showed protective behavior and stable oxide growth throughout the test. The results of the examination of the oxide layers with microbeam X-ray diffraction show clear differences in the structure of protective and nonprotective oxides both at the oxide-metal interface and in the bulk of the oxide layer. The nonprotective oxide interfaces show a smooth transition from metal to oxide with metal diffraction peaks disappearing as the monoclinic oxide peaks appear. In contrast, the protective oxides showed a complex structure near the oxide-metal interface, showing peaks from Zr3O suboxide and a highly oriented tetragonal oxide phase with specific orientation relationships with the monoclinic oxide and the base metal. The same interfacial structures are observed through their diffraction signals in protective oxide layers formed during both 360°C and 500°C corrosion testing. These diffraction peaks showed much higher intensities in the samples from 500°C testing. The results for the various model alloys are discussed to help elucidate the role of individual alloying elements in oxide formation and the influence of oxide microstructure on the corrosion mechanism.
Download or read book Microstructure and Growth Mechanism of Oxide Layers Formed on Zr Alloys Studied with Micro Beam Synchrotron Radiation written by A. Yilmazbayhan and published by . This book was released on 2005 with total page 28 pages. Available in PDF, EPUB and Kindle. Book excerpt: The structures of oxides formed in water and lithiated water on three Zr-based alloys with varied corrosion behavior were studied with micro-beam synchrotron radiation and optical microscopy. Micro-beam synchrotron radiation (0.2 ?m spot) has a unique combination of high elemental sensitivity (ppm level) and fine spatial resolution that allowed the determination of various oxide characteristics such as phase content, texture, grain size, and composition as a function of distance from the oxide-metal interface.
Download or read book Advanced Synchrotron Radiation Techniques for Nanostructured Materials written by Chiara Battocchio and published by MDPI. This book was released on 2019-10-29 with total page 138 pages. Available in PDF, EPUB and Kindle. Book excerpt: Nanostructured materials exploit physical phenomena and mechanisms that cannot be derived by simply scaling down the associated bulk structures and phenomena; furthermore, new quantum effects come into play in nanosystems. The exploitation of these emerging nanoscale interactions prompts the innovative design of nanomaterials. Understanding the behavior of materials on all length scales—from the nanostructure up to the macroscopic response—is a critical challenge for materials science. Modern analytical technologies based on synchrotron radiation (SR) allow for the non-destructive investigation of the chemical, electronic, and magnetic structure of materials in any environment. SR facilities have developed revolutionary new ideas and experimental setups for characterizing nanomaterials, involving spectroscopy, diffraction, scatterings, microscopy, tomography, and all kinds of highly sophisticated combinations of such investigation techniques. This book is a collection of contributions addressing several aspects of synchrotron radiation as applied to the investigation of chemical, electronic, and magnetic structure of nanostructured materials. The results reported here provide not only an interesting and multidisciplinary overview of the chemicophysical investigations of nanostructured materials carried out by state-of-the-art SR-induced techniques, but also an exciting glance into the future perspectives of nanomaterial characterization methods.
Download or read book Microstructure of the Oxide Films Formed on Zirconium Based Alloys written by GP. Airey and published by . This book was released on 1974 with total page 14 pages. Available in PDF, EPUB and Kindle. Book excerpt: Microstructural characterization of oxide films formed on zirconium-based alloys was performed by use of scanning and transmission electron microscopy. Examination of pre-transition films formed on Zircaloy-4 oxidized in 360°C (680°F) water revealed a small grain size of approximately 100 Å (10 nm) diameter. In addition, a gradation of grain size was observed throughout the film thickness, such that at the oxide-water surface (oldest oxide), a grain size of less than 50 Å (5nm) was observed, and at the metal-oxide surface (newest oxide), the grain size was approximately 200 Å (20 nm). In post-transition films the outermost oxide still possessed the very fine 50 Å (5 nm) diameter grain size. However, the newest oxide of post-transition films consisted of relatively large grains, with grain diameters of 1000 to 5000 Å (100 to 500 nm). At the midthicknesses of these oxides intermediate grain sizes were observed. The bulk of the post-transition films was highly porous. Pore sizes ranged from approximately 10 to 150 Å (1 to 15 nm), and many connected pores were concentrated at the grain boundaries. Under more severe oxidizing conditions, imposed by increasing the corrosion temperature to 427°C (800°F), the growth of large grains at the metal-oxide interface was unstable and film growth proceeded by the nucleation of finer grains.
Download or read book Comprehensive Nuclear Materials written by and published by Elsevier. This book was released on 2020-07-22 with total page 4871 pages. Available in PDF, EPUB and Kindle. Book excerpt: Materials in a nuclear environment are exposed to extreme conditions of radiation, temperature and/or corrosion, and in many cases the combination of these makes the material behavior very different from conventional materials. This is evident for the four major technological challenges the nuclear technology domain is facing currently: (i) long-term operation of existing Generation II nuclear power plants, (ii) the design of the next generation reactors (Generation IV), (iii) the construction of the ITER fusion reactor in Cadarache (France), (iv) and the intermediate and final disposal of nuclear waste. In order to address these challenges, engineers and designers need to know the properties of a wide variety of materials under these conditions and to understand the underlying processes affecting changes in their behavior, in order to assess their performance and to determine the limits of operation. Comprehensive Nuclear Materials, Second Edition, Seven Volume Set provides broad ranging, validated summaries of all the major topics in the field of nuclear material research for fission as well as fusion reactor systems. Attention is given to the fundamental scientific aspects of nuclear materials: fuel and structural materials for fission reactors, waste materials, and materials for fusion reactors. The articles are written at a level that allows undergraduate students to understand the material, while providing active researchers with a ready reference resource of information. Most of the chapters from the first Edition have been revised and updated and a significant number of new topics are covered in completely new material. During the ten years between the two editions, the challenge for applications of nuclear materials has been significantly impacted by world events, public awareness, and technological innovation. Materials play a key role as enablers of new technologies, and we trust that this new edition of Comprehensive Nuclear Materials has captured the key recent developments. Critically reviews the major classes and functions of materials, supporting the selection, assessment, validation and engineering of materials in extreme nuclear environments Comprehensive resource for up-to-date and authoritative information which is not always available elsewhere, even in journals Provides an in-depth treatment of materials modeling and simulation, with a specific focus on nuclear issues Serves as an excellent entry point for students and researchers new to the field
Download or read book Effect of Water Chemistry and Composition on Microstructural Evolution of Oxide on Zr Alloys written by B. X. Zhou and published by . This book was released on 2008 with total page 24 pages. Available in PDF, EPUB and Kindle. Book excerpt: The microstructure of oxide films formed on Zircaloy-4 and Alloy No. 3, which has a composition similar to ZIRLOTM, was investigated by high resolution transmission and scanning electron microscopy, and by scanning probe microscopy after corrosion tests performed at 360°C/18.6 MPa in deionized water or lithiated water with 0.01 M LiOH. The microstructural evolution of the oxide films was analyzed by comparing the microstructure at different depths in the oxide layer. The defects, consisting of vacancies and interstitials, such as points, lines, planes, and volumes, were produced during the oxide growth. Monoclinic, tetragonal, cubic, and amorphous phases were detected and their coherent relationships were identified. The characteristic of oxide with such microstructure had an internal cause, and the temperature and time were the external causes that induced the microstructural evolution during the corrosion process. The diffusion, annihilation, and condensation of vacancies and interstitials under the action of stress, temperature, and time caused stress relaxation and phase transformation. It was observed, in the middle of the oxide layer, that the vacancies absorbed by grain boundaries formed pores to weaken the bonding strength between grains. Pores formed under compressive stress lined up along the direction parallel to the compressive stress. Thus, cracks developed from the pores were parallel to the oxide/metal interface. Li+ and OH- incorporated in oxide films were adsorbed on the wall of pores or entered into vacancies to reduce the surface free energy of the zirconium oxide during exposure in lithiated water. As a result, the diffusion of vacancies and the formation of pores were enhanced, inducing the degradation of the corrosion resistance. The relationship between the corrosion resistance of zirconium alloys and the microstructural evolution of oxide films affected by water chemistry and composition is also discussed.
Download or read book The Nature of Unstable Oxide Growth in Zirconium and Zirconium Alloys written by Brendan Ensor and published by . This book was released on 2016 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Zirconium alloys are commonly used as fuel claddings in nuclear reactors due in part to theirsuperior corrosion resistance. The addition of small concentrations of alloying elements prevents thecladding material from undergoing unstable oxide growth under the operating conditions of a nuclearreactor. Unstable oxide growth can also occur due to the presence of hydrides or exposure to neutron flux.The role of alloying elements in avoiding the transition from stable to unstable growth is examined in thisthesis. The goal is to determine the mechanism whereby oxide stabilization occurs.To accomplish this goal, a variety of experiments were performed, and the resulting oxide layerscharacterized with various techniques. Ten model Zr alloys were fabricated and tested in furnace at 600Cfor 40 hours in oxygen and in autoclave at 360C for up to 70 days to determine the causes of breakawayoxidation in pure Zr (and Zr alloys with small concentrations of alloying elements) and the role that alloyingelements play in causing this phenomenon. These alloys were carefully selected and included crystal barZr, sponge Zr, and alloys with small concentrations of Sn, Fe, and Cr. After testing, the alloys werecharacterized using scanning electron microscopy (SEM), Raman spectroscopy, and synchrotron -X-rayfluorescence (XRF) to determine how the structure of the oxide, tetragonal phase content, and alloyingelement distribution affected the formation of unstable oxide. Heterogeneous distribution of alloyingelements was linked to regions of unstable oxide (either nodule-like, grain boundary penetration, ordifferential grain-to-grain growth) and hypothesized to cause breakaway corrosion.The examination of stable oxide layers was then used as a baseline for comparison to cases ofunstable oxide growth in Zr and Zr alloys. One of the primary modes of examination of stable oxide layersformed on Zr alloys was microbeam synchrotron X-ray radiation diffraction and fluorescence, performedat the Advanced Photon Source (APS) at Argonne National Laboratory. This synchrotron X-ray source wasused to perform -X-ray diffraction (XRD), XRF, and 3D Laue spectroscopy. The XRD technique wasused to determine the oxide layer phase content, strain, and grain size as a function of corrosion temperatureand oxide thickness. The XRF technique was used to perform Fe X-ray absorption near-edge spectroscopyiv(XANES) to determine the oxidation state of Fe in the metal as a function of distance from the metal-oxideinterface for various corrosion temperatures. The 3D Laue spectroscopy technique was used to determineplastic deformation and elastic strain in the metal as a function of distance from the metal-oxide interface,corrosion temperature, and oxide thickness for crystal bar Zr and Zircaloy-4.Additionally, Zircaloy-4 samples were corroded in autoclave at 360C for up to 2804 days in andwere periodically weighed to determine oxide thickness. These samples had different coupon thicknessesthat altered the surface-to-volume ratio and led to a higher concentration of hydrogen for a given amountof oxide layer growth. The concentration of hydrogen was measured in archived samples to determine theeffect of hydrogen concentration on corrosion rate. It was observed that the corrosion rate of Zircaloy-4increased with increasing hydrogen concentration above the terminal solid solubility (TSS) of the material(and thus the precipitation of hydrides). More hydrogen caused earlier kinetic transition and areas ofadvanced oxide growth were associated with the locations of hydrides in the metal. It was hypothesizedthat the hydrides hardened the metal ahead of the interface and that the metal was then less able toaccommodate oxide growth stresses leading to earlier kinetic transition and mechanical cracking of theoxide.Finally, eleven Zircaloy-4 samples exposed to various temperatures (272-355C) and neutron fluxlevels (0-11.48 x 1013 n/cm2/s, E > 1 MeV) were examined using XRD and XRF to determine the effectof irradiation on oxide grain size, phase content, and the oxidation of Fe at the APS. With increasing neutronfluence, the grain size of the oxide increased, leading to less tetragonal phase in the oxide away from themetal-oxide interface. At the metal-oxide interface, higher amounts of tetragonal phase were observed withincreasing neutron fluence. This could be caused by the redistribution of Fe from second phase particles(SPPs) into the matrix or due to the hardening of the Zr matrix caused by the exposure to neutrons.The cases of unstable oxide growth examined here were linked to both the distribution and presenceof alloying elements in Zr and Zr alloys and to the level of stress in the oxide. These two phenomena appearto be the primary causes leading to regions of advanced oxide growth and careful consideration should begiven to them when designing and using future Zr alloys in advanced nuclear reactor concepts.
Download or read book Microstructure of Oxides on Zircaloy 4 1 0Nb Zircaloy 4 and Zircaloy 2 Formed in 10 3 MPa Steam at 673 K written by H. Anada and published by . This book was released on 1996 with total page 20 pages. Available in PDF, EPUB and Kindle. Book excerpt: The microstructure of ZrO2 formed on sheet materials of Zircaloy-2 (Zr2), Zircaloy-4 (Zr4), and an alloy of 1.0% Nb added to Zircaloy-4 (1Nb-Zr4) was analyzed using HRTEM (high-resolution transmission electron microscopy). The relationship between the corrosion behavior of the alloys and the microstructure is discussed. Stress-relieved sheet specimens of the three alloys were prepared and corrosion tested under static conditions in steam at 673 K and 10.3 MPa for a total of 220 days. The order of corrosion resistance in 673-K steam was Zr2, 1Nb-Zr4, and Zr4. Several transitions were observed in the corrosion kinetic curve of 1Nb-Zr4 and Zr2. However, only the first transition was observed in the curve of Zr4. Oxide structure in the pre-transition region on Zr4 was analyzed to be in the following order from the outside surface: columnar m-ZrO2, t-ZrO2 layer, substoichiometric Zr oxide layer, and ?-Zr matrix. The t-ZrO2 layer was approximately 50 to 80 nm thick, and the substoichiometric Zr oxide layer was approximately 100 to 200 nm. These layers were absent in the microstructure of the oxide in the post-transition region. The substoichiometric Zr oxide layer consisted of m-ZrO2 grains that were less than 10 nm in diameter and some as yet unidentified grains that had lattice parameters similar to distorted and significantly oriented ?-Zr. However, the t-ZrO2 layered structure and the substoichiometric Zr oxide layer structure were observed in the post-transition oxides on Zr2 and 1Nb-Zr4. It was also observed that transformation of columnar grains to fine equiaxed grains had occurred near the lateral cracks and the incorporated intermetallic precipitates in post-transition oxides. It is implied from these results that the t-ZrO2 layer and the substoichiometric Zr oxide layer structures play an important role as a barrier layer in controlling the occurrence of kinetic transitions.
Download or read book Microstructure Evolution in Zr Alloys During Irradiation written by M. Griffiths and published by . This book was released on 2010 with total page 10 pages. Available in PDF, EPUB and Kindle. Book excerpt: The performance of zirconium alloys in BWR, PWR, and PHWR nuclear reactors is dependent on the microstructure. Accordingly, the characterization of the microstructure is an integral part of any study conducted to develop models for in-reactor performance. Although the as-fabricated microstructure (texture, grain size, dislocation density, and phase or precipitate distribution) determines the basic physical properties of a given component, there are changes that occur during irradiation that can have a significant effect on these properties. Microstructures that illustrate specific features of the radiation damage that forms in Zr alloys will be illustrated and discussed in terms of the dose, dose rate, and impurity factors that are applicable. The original paper was published by ASTM International in the Journal of ASTM International, December 2007.
Download or read book Microstructure Evolution in Ion Irradiated Oxidized Zircaloy 4 Studied with Synchrotron Radiation Microdiffraction and Transmission Electron Microscopy written by Kimberly Colas and published by . This book was released on 2018 with total page 30 pages. Available in PDF, EPUB and Kindle. Book excerpt: The corrosion process (oxidation and hydriding) of zirconium alloy fuel cladding is one of the limiting factors on fuel rod lifetime, particularly for Zircaloy-4. The corrosion rate of this alloy shows indeed a great acceleration at high burnup in light water reactors (LWRs). Understanding the corrosion behavior under irradiation for this alloy is an important technological issue for the safety and efficiency of LWRs. In particular, understanding the effect of irradiation on the metal and oxide layers is a key parameter in the study of corrosion behavior of zirconium alloys. In this study, Zircaloy-4 samples underwent helium and proton ion irradiation up to 0.3 dpa, forming a uniform defect distribution up to 1 ?m deep. Both as-received and precorroded samples were irradiated to compare the effect of metal irradiation to that of oxide layer irradiation. After irradiation, samples were corroded to study the impact of irradiation defects in the metal and in preexisting oxide layers on the formation of new oxide layers. Synchrotron X-ray microdiffraction and microfluorescence were used to follow the evolution of oxide crystallographic phases, texture, and stoichiometry both in the metal and in the oxide. In particular, the tetragonal oxide phase fraction, which has been known to play an important role in corrosion behavior, was mapped in both unirradiated and irradiated metals at the submicron scale and appeared to be significantly affected by irradiation. These observations, complemented with electron microscopy analyses on samples in carefully chosen areas of interest, were combined to fully characterize changes caused by irradiation in metal and oxide phases of both alloys.
Download or read book Zirconium in the Nuclear Industry written by Gerry D. Moan and published by ASTM International. This book was released on 2002 with total page 891 pages. Available in PDF, EPUB and Kindle. Book excerpt: Annotation The 41 papers of this proceedings volume were first presented at the 13th symposium on Zirconium in the Nuclear Industry held in Annecy, France in June of 2001. Many of the papers are devoted to material related issues, corrosion and hydriding behavior, in-reactor studies, and the behavior and properties of Zr alloys used in storing spent fuel. Some papers report on studies of second phase particles, irradiation creep and growth, and material performance during loss of coolant and reactivity initiated accidents. Annotation copyrighted by Book News, Inc., Portland, OR.
Download or read book Correlation Between Characteristics of Oxide Films Formed on Zr Alloys in BWRs and Corrosion Performance written by S. Shimada and published by . This book was released on 2000 with total page 21 pages. Available in PDF, EPUB and Kindle. Book excerpt: The electric resistance of oxide films formed on various Zr alloys in steam at 673 K and in BWRs were measured by an a-c impedance method to investigate the mobilities of the electric charges. The effects of oxide microstructure on transport of electric charges were investigated by TEM.
Download or read book Growth and Characterization of Oxide Films on Zirconium Niobium Alloys written by PK. Chan and published by . This book was released on 1994 with total page 17 pages. Available in PDF, EPUB and Kindle. Book excerpt: Pressure tubes for CANDU reactors are made from extruded and cold-drawn Zr-2.5Nb alloy. Their microstructure consists of elongated ?-Zr grains (0.3 to 0.5 ?m thick), containing about 1 atom percent niobium, surrounded by a thin (30 to 50 nm) network of metastable ?-Zr phase, containing about 20 atom percent niobium. Alloys of Zr-1Nb and Zr-20Nb were prepared, heat treated, and oxidized in 573 K water to produce bulk microstructures and oxides that would simulate those normally found on a much finer scale in pressure tubes. These were subsequently characterized by chemical analyses, scanning electron microscopy (SEM), X-ray diffraction (XRD), analytical electron microscopy (AEM), secondary ion mass spectroscopy (SIMS), X-ray photoelectron spectroscopy (XPS), and nuclear reaction analyses (NRA).
Download or read book Texture and Anisotropy written by U. F. Kocks and published by Cambridge University Press. This book was released on 2000-08-15 with total page 672 pages. Available in PDF, EPUB and Kindle. Book excerpt: A successful book covering an important area of materials science, now available in paperback.
Download or read book Microstructure of Oxide Layers Formed During Autoclave Testing of Zirconium Alloys written by H-O Andrén and published by . This book was released on 1994 with total page 20 pages. Available in PDF, EPUB and Kindle. Book excerpt: The microstructure of oxide layers formed in steam in a 400°C, 10.3-MPa autoclave on different zirconium alloys was studied by transmission electron microscopy. Pre-and post-transition oxide layers on Zircaloy-4 with different heat treatments, and post-transition oxide layers on Zr-0.5Sn-0.53Nb were compared. Special attention was paid to the oxide-metal interface. In Zircaloy-4 with short annealing times and high post-transition corrosion rates, the interface had a disordered structure, and pores were found in the oxide very close to the interface. In Zircaloy-4 with low uniform corrosion rates, the interface consisted of highly ordered, columnar grains. The interface in Zr-0.5Sn-0.53Nb had a different appearance, with an intermediate phase of equiaxed grains between the columnar oxide and the metal. The hydrogen absorption of the zirconium alloys during oxidation was measured by the melt extraction technique on samples oxidized for 63, 147, and 343 days. The Zr-0.5Sn0.53Nb alloy had considerably lower hydrogen absorption than Zircaloy-4.
Download or read book Structure of Oxide Layers Formed on Candidate Steel Alloys Exposed to Flowing Lead bismuth Eutectic for Generation IV Reactor Applications written by Jamie Kunkle and published by . This book was released on 2009 with total page 146 pages. Available in PDF, EPUB and Kindle. Book excerpt: Ferritic-martensitic steels of interest for use in Generation IV lead cooled fast reactors were corroded in a flowing lead-bismuth eutectic environment and the microstructures of the oxide layers they formed were characterized using microbeam synchrotron radiation. Five samples were studied, four of which (HT-9, HT-9 Annealed, T91, and a model alloy) were corroded at 500°C for 666 hours and one of which (HT-9) was corroded at 550°C for 3000 hours in flowing lead bismuth eutectic environments. Studies performed on oxide layers using microbeam synchrotron radiation yielded a detailed view of fluorescence and diffraction data from each of the oxide layers and sublayers formed. Each alloy exhibited a duplex oxide structure consisting of an inner and outer oxide layer. The interface of these two layers corresponded to the original pre-corrosion metal surface. In general, the oxide layers appeared to have been formed in a manner similar to those formed in other gaseous and liquid environments i.e. via the simultaneous ingress of oxygen (O2- ) and egress of iron (Fe2+) across the inner oxide - outer oxide interface. The outer oxide layers observed were formed entirely of Fe3O4 magnetite, contained contaminates (Pb, Bi) from the coolant, and showed evidence of liquid metal dissolution. Inner oxide layers were formed primarily from Fe3O4, but also contained retained ferrite from the bulk metal, carbides, and chromium oxides. Chromium oxides, specifically Cr2O3, are known to act as barriers against the diffusion of oxygen and iron in these materials, slowing oxidation in these materials. The retained bcc iron, or ferrite, in the inner oxide combined with preferential oxidation along lath boundaries suggest that the oxide front advancement proceeded into the metal via preferential oxidation along lath boundaries, followed by selective oxidation of the laths.
Download or read book Oxide Characteristics and Their Relationship to Hydrogen Uptake in Zirconium Alloys written by BD. Warr and published by . This book was released on 1991 with total page 18 pages. Available in PDF, EPUB and Kindle. Book excerpt: Secondary ion mass spectrometry (SIMS) and transmission electron microscopy (TEM) have been used to investigate composition and structure of oxides on pure zirconium and Zr-2.5Nb following both in and out-reactor exposures in aqueous and gaseous environments. Thin oxides formed in steam at 400°C on Zr-2.5Nb act as excellent hydrogen permeation barriers for CANadian Deuterium Uranium (CANDU) pressure tubes. Following up to 4350 effective full power days (EFPD) exposure in-reactor in the annulus gas, and out-reactor elevated exposures to deuterium gas, these oxides generally continue to show diffusional-type through-thickness deuterium concentration profiles, with negligible deuterium contents at the metal/oxide interface. Diffusion coefficients inferred from these profiles are as low as ~2 x 10-22m2/s at 300°C. The structure of these thin oxides on Zr-2.5Nb consists of columnar grains with amorphous regions at grain boundaries and at the metal oxide interface, and non-interconnected porosity, which implies that deuterium permeation is likely controlled by solid state diffusion through the bulk oxide. At regions containing relatively high deuterium contents in the bulk metal of removed pressure tubes, outside surface oxides showed several regions with flat deuterium concentration profiles with higher deuterium concentrations at the metal/oxide interface. Examination of thicker oxides with interconnected porosity, on pure Zr, following exposures to pure deuterium gas, also showed the presence of flat deuterium concentration profiles. This would tend to suggest that regions of high deuterium concentration dissolved in the base metal of pressure tubes may also contain oxides with interconnected porosity.
