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Abstract: During dry storage, creep is the most likely degradation mechanism for spent Zircaloy fuel cladding. The fuel cladding integrity during dry storage depends on the amount of oxidation, irradiation hardening and hydrogen-uptake during in-reactor operation. In this paper, the effect of hydrogen on the creep behavior of Zircaloy-4 cladding material was investigated at different temperatures. Depending on temperature, hydrogen can be found in the sample in solid solution and/or hydride. To capture this phenomenon, a numerical mesoscale model of the hydrogenated material has been built using the Finite Element (FE) Method. The numerical setup explicitly describes the hydrides as an inclusion in a hydrogenated Zircaloy-4 matrix. The matrix creep behavior follows a combined Norton-Bailey and Norton creep rules whereas the hydrides are considered to be elastic material. The creep law was defined in FE Code ABAQUS using the user subroutine CREEP. The comparison of predicted creep behavior obtained from numerical modeling showed good agreement with the results reported in literature. The predicted creep behavior shows a significant effect of hydrides morphology. Particularly, our model is able to seize the competition between the creep strain rate enhancement induced by hydrogen in solid solution and its reduction due to precipitated hydrogen. [Copyright &y& Elsevier]

Features of sheared edges are predicted based on material properties of Zircaloy obtained from the tensile test and ductile fracture model such as the Gurson-Tvergaard-Needleman (GTN) and Johnson-Cook models. The sheared edges formations are numerically analyzed in each ductile model. An appropriate ductile fracture model is selected to study the relative depth of sheared edges with respect to process parameters. The tendency of failure parameters that are affected by sheared edges and fracture duration is investigated. We applied changes on parameters of failure models to show that the punch force curve and the ratio of characteristic lengths could be coincided, which led us to conclude that the GTN and Johnson-Cook models are equivalent. In the Johnson-Cook model, however, the characteristic length of the sheared edges does not change as each failure parameter reaches a critical value. Hence, the FE prediction model for forming defects is developed using the GTN failure model. Finally, the characteristic length of sheared edges have been measured using the FE prediction model for shearing process parameters such as punch velocities, clearance, and tool wear. Our results showed that the punch-die clearance is the most significant factor that affects forming defects when compared to other factors. [ABSTRACT FROM AUTHOR]

Abstract: Nuclear reactor fuel pins act as barriers to the release of radioactive fission products to the coolant flowing around these thin-walled tubes and hence they prevent the leakage of radioactivity to the surroundings of reactor core. These tubes are of small thickness in order to have less resistance in the path of heat flow from the fuel to the coolant. Investigation of failure behavior of these fuel clad tubes is of utmost importance to the designers and plant operators in order to ensure the maximum residence time of the fuel bundles inside the reactor core as well as to ensure minimal activity during operation and refueling activities. Various types of zirconium based alloys are used to manufacture these pins. The focus is to obtain better strength, ductility, corrosion resistance, oxidation resistance, and minimal creep including those due to irradiation-assisted damage and deformation processes. Two number of such types of alloys, namely, re-crystallization annealed Zircaloy-2 and stress-relief annealed Zircaloy-4, have been investigated in this work for their fracture behavior. As standard fracture mechanics specimens cannot be machined from these thin-walled tubes, non-standard specimens with axial cracks have been used in this work. Load normalization technique has been used to evaluate crack growth during loading of these specimens. It was observed that the re-crystallization annealed Zircaloy-2 specimens have higher initiation fracture toughness as well as higher resistance to crack growth compared to the other type of specimens. In order to understand the micro-structural aspects of the fracture resistance behavior of these materials, further investigation incorporating optical and transmission electron microscopy have also been carried out. It was concluded that the higher fracture resistance behavior of the re-crystallization annealed Zircaloy-2 specimens can be attributed to the presence of finer grain and sub-grain micro-structure, very low dislocation density and other defects in the material. [Copyright &y& Elsevier]

Abstract: The zirconium-based alloy Zircaloy-2 contains small amounts of iron, chromium and nickel dissolved in the matrix. Several attempts to measure these amounts have been made in the past, but the results are conflicting and inconclusive. The advent of wide angle, laser pulsed atom probe tomography motivates a new attempt to analyze the matrix. Large datasets are now easily obtained using laser pulsing but quantification is not straightforward due to rather complex mass spectra. Zircaloy-2 contains about 1wt% tin, 0.1wt% oxygen and trace amounts of Si, C and Al. Severe overlaps make quantification of any Fe+, Cr+ and Ni+ ions impossible. Quantification of Fe, Cr and Ni therefore requires that they appear as doubly charged ions only, and consequently the field must be kept high enough. In addition, adsorbed CO+ may appear at the main peak of Fe2+. In the paper a method is reported, which gives what we believe an accurate quantitative analysis of at least iron and chromium in the matrix. [Copyright &y& Elsevier]

We have investigated the effects of Sn and Nb on dislocation properties in a Zr lattice to elucidate the role of these alloying elements in hydride nucleation processes. According to experimental observations, γ-hydride habit planes are close to the prismatic plane in pure Zr and close to the basal plane in Zircaloy. Dislocation loops are observed around hydride precipitates, implying they play a part in hydride formation. Our ab initio generalized stacking-fault energy calculations showed remarkable effects of Sn on unstable-stacking energy and stacking-fault energy: these parameters for basal slip were considerably reduced while those for prismatic slip were increased in the presence of Sn. These results suggest selective stabilization and enhancement of dislocation spreading in the basal plane, promoting possible elementary processes of hydride precipitation with basal habit plane, i.e. screw-dislocation spreading and edge-dislocation emission in the basal plane. [ABSTRACT FROM AUTHOR]

Abstract: Advanced analytical techniques have been used to characterize nuclear materials at the Paul Scherrer Institute during the last decade. The analysed materials ranged from reactor pressure vessel (RPV) steels, Zircaloy claddings to fuel samples. The processes studied included copper cluster build up in RPV steels, corrosion, mechanical and irradiation damage behaviour of PWR and BWR cladding materials as well as fuel defect development. The used advanced techniques included muon spin resonance spectroscopy for zirconium alloy defect characterization while fuel element materials were analysed by techniques derived from neutron and X-ray scattering and absorption spectroscopy. [ABSTRACT FROM AUTHOR]

In our energy hungry world of the twenty-first century, the future of electricity generation must meet the twin challenges of security of supply and reduced carbon emissions. The expectations for nuclear power programmes to play a part in delivering success on both counts, grows ever higher. The nuclear industry is poised on a renaissance likely to dwarf the heady days of the 1960s and early 1970s. Global supply chain and project management challenges abound, now just as then. The science and engineering of materials will be key to the successful deployment and operation of a new generation of reactor systems and their associated fuel cycles. Understanding and predicting materials performance will be key to achieving life extension of existing assets and underpinning waste disposal options, as well as giving confidence to the designers, their financial backers and governments across the globe, that the next generation of reactors will deliver their full potential. [ABSTRACT FROM AUTHOR]

Abstract: Hydrogen embrittlement of zirconium alloys is one of the main causes of the mechanical degradation of the fuel cladding in light water reactors, and has therefore been extensively studied. Although various conjectures have been proposed as the origin of such embrittlement, it is not known which one plays the most important role: the brittle nature of the hydride, micro-crack nucleation by interaction of hydride precipitates with dislocations or void nucleation at the interface between hydride precipitates and zirconium matrix. The purpose of the present study was to elucidate the origin of the embrittlement by investigating the fracture properties of the hydride. We have evaluated the surface energy γ S and unstable stacking energy γ US of Zr–H systems by using ab initio calculations. The ductile/brittle behavior of the hydride is discussed based on the difference between γ S and γ US among the hydride, pure zirconium and hydrogen solid solution. For the solid solution at a H/Zr ratio less than 0.5 we obtained a monotonous decrease by 15–34% and 50–100% in γ S and γ US, respectively, from those in pure zirconium, indicating a reduction in both brittleness and ductility. Thus, hydrogen-induced embrittlement of the hcp Zr matrix was not confirmed. On the other hand, for the hydride we obtained a 25% smaller γ S and a 200–300% larger γ US than those in pure zirconium. This indicates that zirconium hydride has an extremely brittle nature due to the synergistic effect of a small γ S, implying easy generation of a fracture surface, and large γ US, implying a difficulty in dislocation motion, compared with pure zirconium. Furthermore, Rice’s ductile/brittle parameter D was 1.4 in the δ-hydride, indicating that it undergoes brittle fracture more easily than iridium, known as an extremely brittle material. These results seem sufficient to attribute hydrogen embrittlement of zirconium alloys substantially to the brittle nature of the hydride. [Copyright &y& Elsevier]

Abstract: This paper gives an overview on the status of knowledge of high-temperature oxidation of the two zirconium alloys Zircaloy-4 and E110 with special emphasis on results obtained during the SARNET period. The tin-bearing alloy Zircaloy-4 and the niobium-bearing alloy E110 are the materials for cladding and structures used in pressurised water reactors of the Western-type and VVERs and RBMKs, respectively. Results of separate-effects tests, single-rod tests, and large-scale bundle experiments are summarised. Focus is directed to oxidation kinetics at high temperature, hydrogen release and absorption by the remaining metal, and behaviour during quenching. Both materials behave very similarly as long as the superficial oxide scales formed during oxidation are dense and protective. Main differences are seen in connection with breakaway oxidation which leads to enhanced oxidation and hydrogen uptake and thus embrittlement and possibly earlier failure of the cladding. The temperature range at which pronounced breakaway is observed is different for the two alloys. The status of modelling of oxidation kinetics, thermo-mechanical behaviour during cooldown and the influence of irradiation are discussed at the end of the paper. [Copyright &y& Elsevier]

The interdiffusion reactions between Zircaloy-2 and a Ni based alloy (Inconel 750X) have been studied in the temperature range 1083-1173 K by employing annealing of diffusion couples. The reaction zone developed at the interface was characterised using an optical microscope and electron probe microanalyser (EPMA). The diffusion reaction between Zircaloy-2 and Inconel shows that the reaction products form both as a layered structure and as two phase mixture in the diffusion zone. Three layers were found to form in the diffusion zone, namely, Zr2Ni (ζ phase) and ψ1 phase close to Zircaloy-2 and a mixture of ψ1 and ψ2 phases close to Inconel. The kinetics for the growth of the three layers as well as the total diffusion zone was established. The growth of all the layers shows Arrhenius type of relation with respect to temperature. The temperature dependence for the reaction constants, K, for the various phases was also established. [ABSTRACT FROM AUTHOR]

Abstract: The assembly of various materials by the conventional methods finds its limits in certain specific applications. We approach a study on the solid state diffusion bonding between zircaloy (Zy4) and stainless steel (304L) for an application in the sector of the nuclear power. The diffusion couples prepared underwent treatments at the temperatures ranging between 850 ∘C and 1020 ∘C in a controlled atmosphere and under dynamic pressures of contact. We give a particular attention to the morphology of the interface, formed, with the chemical reactions of exchange which proceed between two entities Zy4/304L like with the determination of the nature of the compounds formed with the interface. The observations and chemical analysis are realized by optics microscopy’s and electronics (ESEM-EDX) and with X ray diffraction (XRD). The quantitative distributions as well as the detailed localization of the basic chemical elements are defined by chemical profiles, characteristics of each element. The junction with the interface of diffusion consists of three zones distinct, formed from a solid solution FeCr, rich in Cr in the form of a homogeneous edge, localized in steel side. The zone of the center and that of side Zy4 are two-phase. They are made up, for one, of FeCr- Zr(Fe, Cr)2 and for the other, Zy4 side, the solid solution Zr and the intermetallic compound Zr2 (Fe1−x Ni x ) where X is understood enters 0,15 and 0,25. The detailed results obtained, are a stepping, between those obtained from the observations and chemical analyses and radio crystallographic. The values of the measured microhardnesses give a very heterogeneous filiations to the level of the interface of the two entities. [Copyright &y& Elsevier]

The contrastive corrosion experiments between surface nanocrystallined Zircaloy-4 and coarse-grained Zircaloy-4 under the condition of 673 K/10.3 MPa in pure water are carried out, and the microstructure of oxide films has been studied. The results indicate that the growth rate of oxide films formed on the nanocrystalline Zircaloy-4 is lower than that of oxide films formed on the coarse-grained Zircaloy-4. Simultaneously, the oxide/metal interface of the former is more regular and glossy than that of the latter. For nanocrystalline Zircaloy-4, the low oxygen diffusion rate through the oxide/metal interface can hinder the reaction of oxygen ion with metal ion. Furthermore, more tetragonal ZrO2 are observed in the oxide films, which can delay the martensite phase transition from tetragonal to monoclinic phase in oxide films. [ABSTRACT FROM AUTHOR]

A synthesis on the mechanical characteristics of unirradiated and irradiated fuel rod claddings was performed by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) in order to have reference data for the assessment of the safety demonstrations in normal and accident conditions of transport required by the procedure of package licensing. Indeed, the transport conditions correspond to a range of cladding temperatures (200–550°C) which is only partly covered by the data acquired within the framework of the safety demonstration relative to the reactor normal operating conditions, especially beyond 400°C. This work concerned Zircaloy-4 cladding material (Zry-4) and M5TM. Data about mechanical properties (elastic and ductile properties, creep behaviour), oxidation (in reactor and under air during transport), hydrides and fracture toughness have been collected and synthesised. The laws presented in the document can be used to obtain orders of magnitude of oxide layer thickness, hydrogen content and creep deformation rate. The following phenomena which could influence the mechanical behaviour of the cladding were more particularly studied: oxidation which could become very important during transport in case of cladding temperatures of ∼500°C; creep for which only a few data ∼500°C are available and which depends in particular on the internal pressure of the rods, the cladding oxidation and the presence of the hydrides; and recrystallisation of Zry-4 at ∼500°C, which could have consequences on the mechanical properties of the cladding after cooling during the storage. For other topics of interest for the study of the mechanical behaviour of the cladding, such as the fracture toughness for example, it was identified that the data available is scarce. [ABSTRACT FROM AUTHOR]

• The HNGD model was found to be unable to predict the hydride rim thickness under temperature gradient. • Two hypotheses, targeting the supersolubility and solubility, were proposed to address this issue. • The modified HNGD model accurately predicts the thickness of the hydride rim in a sample submitted to a temperature gradient. • The modified HNGD model predicts the hydride distribution in the case of a Zircaloy cladding with an inner Zr liner. • The modified HNGD model has been implemented and verified and is now available for general use in Bison. [Display omitted] During operation of a light water reactor, waterside corrosion of the Zircaloy nuclear fuel cladding causes hydrogen pickup. The absorbed hydrogen can redistribute in the cladding driven by existing concentration, stress, and temperature gradients. When the concentration reaches the solubility limit, hydrides precipitate. These hydrides can be more brittle than the Zircaloy matrix, so they can endanger the cladding integrity during a transient if their concentration is too high. In recent years, extensive efforts have been made to understand hydrogen behavior and to develop simulation tools able to predict hydrogen diffusion and hydride precipitation and dissolution. These efforts led to the development of the Hydride Nucleation-Growth-Dissolution (HNGD) model and its implementation into the nuclear fuel performance code Bison. While it offers a significant improvement and accurately predicts the amount of precipitates, this model fails to predict the thickness of the hydride rim under a temperature gradient. The current work presents the limitation of the HNGD model and proposes two hypotheses to improve the model's accuracy. The first hypothesis introduces a time dependency to the supersolubility to reduce the nucleation barrier as hydrogen atoms find more favorable nucleation sites. The second one introduces a hydride content dependency to the solubility. These hypotheses were validated and implemented into Bison and are now available to the user community. The modified HNGD model accurately predicts the hydride rim thickness, and it was demonstrated that this updated model can be used in Bison to model Zircaloy cladding with a zirconium inner liner. Finally, potential experimental and numerical methods are discussed to further validate these hypotheses. [ABSTRACT FROM AUTHOR]

Abstract: The formation of hydrides in zirconium alloys significantly affects their mechanical properties and is considered to play a critical role in their failure mechanisms, yet relatively little is known about the micromechanical behavior of hydrides in the bulk. This paper presents the result of in situ uniaxial mechanical tensioning experiments on hydrided zircaloy-2 and zircaloy-4 specimens using energy-dispersive synchrotron X-ray diffraction, which suggests that a stress-induced transformation of the δ-hydride to γ-hydride via ordering of the hydrogen atoms occurs, akin to a Snoek-type relaxation. Subsequent annealing was found to reverse the ordering phenomenon. [Copyright &y& Elsevier]

• Developed a mechanistic oxygen distribution model in steam-oxidized Zircaloy-4. • Experimental validation conducted for ECR, oxygen distribution, and phase thickness. • Providing accurate oxidation prediction in non-isothermal conditions. • Validate the accuracy of CP correlation and provide an advanced mechanistic model. • Explaining effect of hydrogen on oxygen content in zr for highburnup cladding. TRANOX-1.0 (TRANsient OXidation), a mechanistic model that calculates oxygen distribution for isothermal and non-isothermal transients (1000–1250 °C) has been developed. TRANOX solves radial transient diffusion equation using Finite Difference Method (FDM) with spatial phase changes for each time step. Diffusion coefficients of each resulting phase were obtained by searching a combination that best guarantees Equivalent Cladding Reacted (ECR) and α-Zr(O) thickness simultaneously. The model has been extensively validated by overarching experimental campaigns including isothermal and non-isothermal high temperature steam oxidation, and post-metallurgical characterization. Accounting for α+β equilibrium phase, TRANOX gives accurate predictions for ECR, spatially averaged radial oxygen distribution, and thicknesses of each phase. Mechanistically solving the diffusion equation, TRANOX exhibits unbiased predictability for isothermal and non-isothermal transients. TRANOX was used to understand the oxygen distribution of pre-hydrided high burnup fuel cladding. The diffusion resistance behavior of α-Zr(O) and α+β with respect to hydrogen content counteract, neutralizing the overall hydrogen effect on oxygen distribution. The limited hydrogen effect on diffusion resistance during the early stage of steam oxidation further reduces the hydrogen-sensitivity of oxygen profiles. This finding provides a cogent explanation for the effect of hydrogen on oxygen distribution and oxidation kinetics observed by a number of past studies. [ABSTRACT FROM AUTHOR]

The oxidation behaviors of Cr, Cr93.4Al6.6, Cr58.1Al41.9, and Cr34.5Al65.5 coatings, deposited by using multi-arc ion plating technology, at high temperature were studied. The weight gain, oxide thickness, morphology, and phase composition of the coatings before and after oxidation were analyzed in detail. The results show that there is an Al content window available for tuning the oxidation behaviors of the CrAl-based coatings. The Cr93.4Al6.6 coating is considered to be most protective and can effectively improve the high-temperature oxidation resistance of the substrate; whereas, too high an Al content has a harmful effect on the antioxidant properties of the coatings. The oxidation mechanism of Cr and CrAl coatings were also discussed. [ABSTRACT FROM AUTHOR]

A computation model of the corrosion rate versus grain size of nanocrystalline zircaloy-4 was presented. The influence of the second phase on the conductivity of alloy was considered. By this model, the corrosion rate of nanocrystalline zircaloy-4 at different temperature was calculated. The results show that the corrosion rate constant and weight gain of nanocrystalline zircaloy-4 decrease with the decrease of grain size, and that the corrosion weight gain of nanocrystalline zircaloy-4 is less than that of zircaloy-4 of coarse grain. The computational result is coincident with the experimental result. [ABSTRACT FROM AUTHOR]

Effect of interface undulation on the high-temperature oxidation behaviors of grit-blasted and coated zircaloys in pressurized water was investigated. The grit-blasted and the coated zircaloy substrates present significantly higher corrosion rate than the as-received flat substrate during the early stage of exposure. Numerical analysis suggests the interface undulation results in redistribution of the oxide growth stress of zircaloy, as well as varying the local stress concentration, which induces initiation of cracks and therefore increases the corrosion rate of zircaloy. Propagation of cracks along the interface drives delamination of the protective coating. Results of this work not only promote understanding of the coating/substrate interface evolution in high-temperature pressurized water, but also provide a guide to design corrosion-resistant zirconium alloys and/or coatings for fuel cladding. • The grit-blasted and coated zircaloys present higher corrosion rate than the as-received substrate during early stage. • Interface undulation can result in redistribution of the oxide growth stress of zircaloy. • Local stress concentration induces initiation of cracks and therefore increases the corrosion rate of zircaloy. [ABSTRACT FROM AUTHOR]

Abstract: A nanocrystalline zircaloy-4 was corroded under 673 K/10.3 MPa, and the change of the lattice constant of ZrO2 during the phase transition was studied. The results show that the change of the lattice constant of ZrO2 formed on the nanocrystalline base is very different from that of lattice constant of ZrO2 formed on the coarse grain base. The unit cell volume of the tetragonal ZrO2 and monoclinic ZrO2 formed on the nanocrystalline base is constricted during the oxidation process. However, for the coarse grain base, the unit cell volume of the tetragonal ZrO2 is constricted, and the unit cell volume of the monoclinic ZrO2 is inflated. This phenomenon was discussed. [Copyright &y& Elsevier]

Abstract: Secondary phase precipitates of a Zircaloy sample have been characterised by X-ray microspectroscopy. In Zircaloy-2 X-ray microscopy reveals pictures with a 40nm resolution identifying Fe, Cr and to a lower occurrence Ni phases up to size of the micrometer. Analysis by X-ray spectroscopy defines the structure of specific secondary phase precipitates. The feasibility tests demonstrate that the characterisation of Fe and Cr can be performed on 100nm size phases allowing the analysis of the Fe or Cr atoms environment in these secondary phase precipitates. [Copyright &y& Elsevier]

Abstract: Hydrogen absorbed in steam-oxidised zircaloy can be determined quantitatively by means of neutron radiography. Correlation parameters between the total cross section and hydrogen content as well as oxide layer thickness were determined quantitatively. At H/Zr atomic ratios lower than 1.0, linear correlations between the hydrogen content and total cross section exist. The total cross section of Zr is lower and the effect of the hydrogen is higher in radiography measurements with a cold neutron spectrum than with a thermal spectrum. A Be filter reduces the effects of lower wavelength and epithermal neutrons and extends the linear correlations to higher H/Zr atomic ratios. Due to the better possibilities of background corrections, the neutron image should be detected by a CCD camera for a proper quantitative analysis with a medium spatial resolution of about 0.1mm. A higher spatial resolution, but larger uncertainties in the quantitative hydrogen determination are achieved by measurements with imaging plates. The effect of oxygen layers on the total cross section is much smaller than the effect of hydrogen. The total cross section measured depends linearly on the oxide layer thickness. [Copyright &y& Elsevier]

Abstract: We present an assessment of the mechanical constitutive behavior after irradiation for a tempered martensitic steel, Eurofer97’, and for Zircaloy by means of small-ball punch tests. For that purpose we use a finite-element model of the punch test device. We take advantage of the model to infer the alloy mechanical constitutive behavior after irradiation. We show that a necessary matching between the experimental and calculated force–deflection curves allows choosing consistently among the proposed constitutive behaviors. We discuss also the limitations of this method. For Zircaloy a complete description of the constitutive behavior in the unirradiated or irradiated condition was not possible in the framework of anisotropic plastic potentials and associated flow rules, since this demands to fully characterize experimentally the anisotropic yield surface and its evolution with plastic strain during punch deformation. However, an approximate description of the initial yielding anisotropy using Hill plastic potential is possible (although not unique) and necessary before an assessment of the irradiation-hardening can be done on a safe basis. Developing a conventional (non-axisymmetric) three-dimensional model was mandatory in order to deal with the problem of anisotropy in Zircaloy. [Copyright &y& Elsevier]

Abstract: This article presents an overview of the surface chemistry of zirconium, focusing on the relationship of what is known from model studies and how this connects to current and future applications of Zr-based materials. The discussion includes the synergistic nature of adsorbate interactions in this system, the role of impurities and alloying elements, and temperature-dependent surface–subsurface transport. Finally, some potential uses of zirconium and its alloys for biomedical and nanolithographic applications are presented. [Copyright &y& Elsevier]

Abstract: Pre-oxide films were formed on the surface of Zircaloy specimens at high temperatures 700 and 1100°C and their corrosion characteristics were investigated at 360°C. The pre-oxides of different thickness were layered with varying annealing times; 0.5–2.3μm thick films were formed by annealing at 700°C from 10 to 120s while oxide layers thinner than 2μm were not obtained at 1100°C even with the shortest exposure time 20s. Cross-sectional studies revealed that the pre-oxides were adherent. Oxidation characteristics of the specimens having pre-oxide layers ranging from 0.5 to 3.7μm were investigated by measuring weight gain in an autoclave at 360°C. The specimens with 1–2.3μm pre-oxide layers exhibited better oxidation resistance than the thinner or thicker oxides. The bare Zircaloy specimen without a pre-oxide layer had the highest weight gain due to the fact that the bare specimen experienced rate transition after 98 days of exposure. Microstructural study showed the possibility that the tetragonal fraction of the pre-oxide affected the weight gain incremental behavior. [Copyright &y& Elsevier]

The understanding of the mechanisms of high temperature growth of zirconium oxide on zircaloy alloys is of considerable importance owing to the world-wide use of zirconium in nuclear plants. Spallations and growth recovery are known to originate in structural changes of the oxide, but the induced modifications in the oxygen transport have still to be investigated. The zirconia is known to grow mainly by anionic transport, but, due to the very high solubility of oxygen in zirconium, the oxygen penetration profile may be rather complex and extend beyond the oxide/metal interface. The nuclear microprobe offers an attractive way to investigate the oxygen penetration, both from RBS and NRA. In the case of parallel mechanical loading and high temperature oxidation, the oxygen diffusivity is very sensitive to the generated defects in the oxide scales. The use of alternate 16O/18O atmosphere and subsequent local microanalysis by means of 18O(p,α)15N nuclear reaction allows to investigate the effect of the mechanical loading on the oxygen penetration. [Copyright &y& Elsevier]