Arthur Motta

This study focuses on the precipitation of nanoscale hydrides in polycrystalline zirconium as a first step to predicting the hydride morphology observed experimentally and investigating the mechanisms responsible for hydride reorientation at the mesoscale. A quantitative phase-field model, which includes the elastic anisotropy of the nanoscale zirconium hydride system, is developed to investigate the mechanism of hydride reorientation in which the presence of an applied hoop stress promotes hydride precipitation in grains with basal poles aligned with the circumferential direction. Although still elongated along the basal plane of the hexagonal matrix, nanoscale hydrides growing in grains oriented perpendicular to the applied stress appear radial at the mesoscale. Thus, a preferential hydride precipitation in grains with basal poles aligned parallel to the applied stress could account for mesoscale hydride reorientation. This mechanism is consistent with experimental observations performed in other studies.

Eleven Zircaloy-4 samples were irradiated in the Advanced Test Reactor at a variety of temperatures and neutron flux levels for up to 6.5 years. Subsequently, the coupons were characterized with complementary techniques to understand the mechanisms behind oxide growth as a function of different corrosion environments. Samples were examined using synchrotron X-ray diffraction/fluorescence, traditional X-ray diffraction, focused ion beam/scanning electron microscopy serial sectioning, and three-dimensional reconstruction to develop an improved understanding of the influence of the underlying oxide microstructure on oxide growth. The oxide microstructure formed under irradiation was compared to that in samples corroded in an autoclave to discern the impact of neutron irradiation and temperature on corrosion rate, oxide kinetic transition, irradiation-induced breakaway corrosion, stress development, phase formation, and oxide grain size. The microstructure of the oxide changed with the corrosion temperature, with larger crack spacing (characteristic of kinetic transition) and larger monoclinic oxide grains formed during higher temperature corrosion. The specimens that were exposed to a neutron flux exhibited larger oxide grains and an increase in the fraction of tetragonal phase at the metal-oxide interface (but less tetragonal phase in the bulk oxide) compared to those exposed in autoclave. Data obtained from electron microscopy demonstrated the effect of irradiation and corrosion temperature on oxide morphology. One specimen underwent an irradiated-induced breakaway oxidation that was characterized by a sharp change in the corrosion rate and a decrease in the spacing between adjacent crack layers in the oxide film. Stress is hypothesized to be a key driver in the oxide growth formation, with samples nearer transition having more plastic deformation in the metal and increased elastic strain. These observations lead to a theory of oxide growth on zirconium alloys that attempts to connect and integrate the effects of stress, irradiation, temperature, phase formation, crystal orientation, porosity, and precipitate amorphization.

Recent concern with fuel safety in accident scenarios has motivated research into accident tolerant fuels (ATF), which are defined as fuels that could increase coping time in case of an accident. This study is an attempt to develop an ATF by improving the corrosion performance of nuclear fuel cladding during a high-temperature excursion through the application of a ceramic coating using physical vapor deposition. In this study, ceramic coatings constituted of single-layer and multi-layer TiN/TiAlN coatings with a titanium bond coat layer to improve adhesion were applied onto ZIRLO sheets using cathodic arc physical vapor deposition. The coating architecture and deposition parameters were systematically optimized to achieve good adhesion and corrosion performance, and an initial evaluation was performed for resistance to radiation damage. The coating performance was highly dependent on coating design architecture, and the best coating architecture was found to be that of eight-layer TiN/TiAlN coatings deposited with optimized parameters. The optimized coatings were corrosion tested in 360°C water for up to 90 days, showing essentially no oxygen penetration, very low weight gain, and no spallation or debonding. The samples were also examined in microscopy and X-ray diffraction after corrosion testing, and little change was observed. To evaluate the coating performance under irradiation, cross-sectional transmission electron microscopy samples of the coating were subjected to in situ ion irradiation to a dose of 20 dpa with 1 MeV Kr ions at 300°C, followed by further annealing to 800°C. Little interlayer mixing and overall damage accumulation was observed. Coating adhesion was investigated through scratch testing and post-scratched sample failure mode characterization to determine a critical load value for spallation. The coating layers are found to require a high load for debonding and spallation. The results suggest that this optimized coating system is a promising path for developing an ATF.

Hydride precipitation may impact the integrity of zirconium-based nuclear fuel cladding, both during normal operation and during extended dry storage. To better understand such degradation, a study of hydride precipitation of zirconium hydrides in Zircaloy-4 samples was performed. The samples were submitted to various thermomechanical cycles using both in situ synchrotron X-ray diffraction and differential scanning calorimetry. Results showed that as the hydrided samples were cooled at moderate to fast cooling rates, the hydrogen content in solid solution (CSS) decreased, following the terminal solid solubility for precipitation (TSSP) curve, reflecting hydride precipitation in the matrix. However, when the samples were held for an isothermal anneal at a fixed temperature, the CSS continued to decrease below TSSP and approached the terminal solid solubility for dissolution (TSSD). This result suggests that TSSP is a kinetic limit and that a unique solubility limit TSSD governs zirconium hydride precipitation. Hydride precipitation rate and the degree of precipitation reaction completion between 280 and 350°C were obtained using differential scanning calorimetry. Using this data, a temperature-time transformation diagram for hydride precipitation in Zircaloy-4 was generated that showed that hydride precipitation is diffusion-driven under 310°C and reaction-driven above 310°C. The experimental data were fitted to the Johnson-Mehl-Avrami-Kolmogorov model and an Avrami parameter of 2.56 was obtained (2.5 is the theoretical value for the growth of platelets). Results imply that hydride nucleation occurs if CSS is greater than TSSP while hydride growth occurs if preexisting hydride platelets are present and CSS is above TSSD. Combined with existing theory, these data were used to develop the hydride growth, nucleation, and dissolution model that can simulate hydrogen behavior in Zircaloy.

Hydride reorientation can occur as a result of vacuum drying or transportation of spent nuclear fuel rods prior to dry cask storage. The elevated temperatures generate high internal gas pressure in the fuel rods, causing δ-hydride platelets to precipitate perpendicular to the hoop stress during cooling. Because the loading causes multiaxial stresses, it is of interest to elucidate the role of stress state on the threshold stress for hydride reorientation. To that end, specially designed specimens were used with a range of stress biaxiality ratios (σ1/σ2) from uniaxial tension (σ1/σ2 = 0) to near-equibiaxial tension (σ1/σ2 = 0.8). The threshold stress was determined in each case by matching the major and minor stresses (and thus the local stress state) calculated by finite-element analysis to the hydride microstructures created by the thermomechanical treatment at that specific location. Using cold-worked stress-relieved Zircaloy-4, the results show that as the stress biaxiality ratio increased from uniaxial tension to near-equibiaxial tension, the threshold stress decreased from 155 to 75 MPa. To elucidate the hydride reorientation process, hydride precipitation and d-spacing behavior were investigated in situ using synchrotron radiation diffraction. The precipitation temperature for out-of-plane hydrides was lower than that for in-plane hydrides. The δ{111} d-spacing aligned with the hydride platelet face was greater than the d-spacing of planes aligned with platelet edges. Furthermore, δ{111} planes exhibited bilinear thermally induced expansion, but only for those planes aligned with hydride plate edges. In contrast, the hydride platelet face contracted upon heating. The experimental results were explained by a reversal of stress state associated with precipitating or dissolving hydrides within α-zirconium. In addition, irradiated cladding after thermomechanical treatments was examined by synchrotron radiation diffraction at ambient temperatures. Although the hydride intensity was low for accurately determining d-spacing, the diffraction patterns indicated that β-niobium peaks present in the un-irradiated cladding were diminished after irradiation.

Because hydrogen ingress into zirconium cladding can cause embrittlement and limit cladding lifetime, hydrogen pickup during corrosion is a critical life-limiting degradation mechanism for nuclear fuel. However, mechanistic knowledge of the oxidation and hydrogen pickup mechanisms is still lacking. In an effort to develop such knowledge, we conducted a comprehensive study that included detailed experiments combined with oxidation modeling. We review this set of results conducted on zirconium alloys herein and articulate them into a unified corrosion theoretical framework. First, the hydrogen pickup fraction (fH) was accurately measured for a specific set of alloys specially designed to determine the effects of alloying elements, microstructure, and corrosion kinetics on fH. We observed that fH was not constant and increased until the kinetic transition and decreased at the transition. fH depended on the alloy and was lower for niobium-containing alloys. These results led us to hypothesize that hydrogen pickup during corrosion results from the need to balance the charge during the corrosion reaction such that fH decreases when the rate of electron transport through the protective oxide increases. To assess this hypothesis, two experiments were performed: (1) micro-X-ray absorption near-edge spectroscopy (μ-XANES) to investigate the evolution of the oxidation state of alloying elements when incorporated in the growing oxide and (2) in situ electrochemical impedance spectroscopy (EIS) to measure oxide resistivity as a function of exposure time on different alloys. With the use of these results, we developed an analytical zirconium alloy corrosion model based on the coupling of oxygen vacancies and electron currents. Both modeling and EIS results show that as the oxide electric conductivity decreases the fH increases. These new results support the general hypothesis of charge balance. The model quantitatively and qualitatively predicts the differences observed in oxidation kinetics and hydrogen pickup fraction between different alloys.

A review is presented of work performed in our group over the years in the areas of radiation damage, corrosion, hydrogen pickup, hydriding, and the mechanical behavior of zirconium alloy nuclear fuel cladding with the goal of developing a greater mechanistic understanding of cladding performance in service.

Hydrogen pickup of zirconium-based fuel cladding and structural materials during in-reactor corrosion can degrade fuel components because the ingress of hydrogen can lead to the formation of brittle hydrides. In the boiling water reactor (BWR) environment, Zircaloy-2 fuel cladding and structural components such as water rods and channels can experience accelerated hydrogen pickup, whereas Zircaloy-4 components exposed to similar conditions do not. Because the principal difference between the two alloys is that Zircaloy-2 contains nickel, accelerated hydrogen pickup has been hypothesized to result from the presence of nickel. However, an understanding of the mechanism by which this acceleration occurs is still lacking. We investigated the link between hydrogen pickup and the oxidation behavior of alloying elements when incorporated into the oxide layers formed on zirconium alloys when corroded in the reactor. Synchrotron radiation microbeam X-ray absorption near-edge spectroscopy (XANES) at the Advanced Photon Source was performed on carefully selected BWR-corroded Zircaloy-2 water rods at an assembly-averaged burnup ranging from 32.8 to 74.6 GWd/MTU to determine the oxidation states of alloying elements, such as iron and nickel, within the oxide layers as a function of distance from the oxide-metal interface at high burnup. Samples were chosen for comparison based on having similar oxide thicknesses, processing, elevation, reactors, and fluences but different hydrogen pickup fractions. Examinations of the oxide layers formed on these samples showed that (1) the oxidation states of these alloying elements changed with distance from the oxide-metal interface, (2) these elements exhibited delayed oxidation relative to the host zirconium, and (3) nickel in Zircaloy-2 remained metallic in the oxide layer at a longer distance from the oxide-metal interface than iron. An analysis of these results showed an apparent correlation between the delayed oxidation of nickel and higher hydrogen pickup of Zircaloy-2 at high burnup.

Extended dry storage of spent nuclear fuel makes it desirable to assess the structural integrity of the storage canisters. Stress corrosion cracking of the stainless steel canister is a potential degradation mode especially in marine environments. Sensing technologies are being developed with the aim of detecting the presence of chloride-bearing salts on the surface of the canister as well as whether cracks exist. Laser-induced breakdown spectroscopy (LIBS) methods for the detection of Chlorine are presented. In addition, ultrasonic-guided wave detection of crack-like notches oriented either parallel or perpendicular to the shear horizontal wave vector is demonstrated using the pulse-echo mode, which greatly simplifies the robotic delivery of the noncontact electromagnetic acoustic transducers (EMATs). Robotic delivery of both EMATs and the LIBS system is necessary due to the high temperature and radiation environment inside the cask where the measurements need to be made. Furthermore, the space to make the measurements is very constrained and maneuverability is confined by the geometry of the storage cask. In fact, a large portion of the canister surface is inaccessible due to the presence of guide channels on the inside of the cask's overpack, which is strong motivation for using guided waves for crack detection. Among the design requirements for the robotic system are to localize and track where sensor measurements are made to enable return to those locations, to avoid wedging or jamming of the robot, and to tolerate high temperatures and radiation levels.

Extended dry storage of spent nuclear fuel makes it desirable to assess the structural integrity of the storage canisters. Stress corrosion cracking of the stainless steel canister is a potential degradation mode especially in marine environments. Sensing technologies are being developed with the aim of detecting the presence of chloride-bearing salts on the surface of the canister as well as whether cracks exist. Laser induced breakdown spectroscopy (LIBS) methods for the detection of Chlorine are presented. Detection of a notch oriented either parallel or perpendicular to the shear horizontal wave vector is demonstrated using the pulse-echo mode, which greatly simplifies the robotic delivery of the noncontact electromagnetic acoustic transducers (EMATs). Robotic delivery of both EMATs and the LIBS system is necessary due to the high temperature and radiation environment inside the cask where the measurements need to be made. Furthermore, the space to make the measurement is very constrained and maneuverability is confined by the geometry of the storage cask. In fact, a large portion of the canister surface is inaccessible due to the presence of guide channels on the inside of the cask’s overpack, which is strong motivation for using guided waves for crack detection. Among the design requirements for the robotic system are: to localize and track where sensor measurements are made to enable return to those locations, to avoid wedging or jamming of the robot, and to tolerate high temperatures and radiation levels.

A new in situ sample environment has been designed and developed to study the interfacial interactions of nuclear cladding alloys with high temperature steam. The sample environment is particularly optimized for synchrotron X-ray diffraction studies for in situ structural analysis. The sample environment is highly corrosion resistant and can be readily adapted for steam environments. The in situ sample environment design complies with G2 ASTM standards for studying corrosion in zirconium and its alloys and offers remote temperature and pressure monitoring during the in situ data collection. The use of the in situ sample environment is exemplified by monitoring the oxidation of metallic zirconium during exposure to steam at 350 °C. The in situ sample environment provides a powerful tool for fundamental understanding of corrosion mechanisms by elucidating the substoichiometric oxide phases formed during the early stages of corrosion, which can provide a better understanding of the oxidation process.

During operation, nuclear fuel rods are immersed in the primary water, causing waterside corrosion and consequent hydrogen ingress. In this review, the mechanisms of corrosion and hydrogen pickup and the role of alloy selection in minimizing both phenomena are considered on the basis of two principal characteristics: the pretransition kinetics and the loss of oxide protectiveness at transition. In zirconium alloys, very small changes in composition or microstructure can cause significant corrosion differences so that corrosion performance is strongly alloy dependent. The alloys show different, but reproducible, subparabolic pretransition kinetics and transition thicknesses. A mechanism for oxide growth and breakup based on a detailed study of the oxide structure can explain these results. Through the use of the recently developed coupled current charge compensation model of corrosion kinetics and hydrogen pickup, the subparabolic kinetics and the hydrogen fraction can be rationalized: Hydrogen pickup increases when electron transport decreases, requiring hydrogen ingress to close the reaction.

Although the optimization of zirconium-based alloys has led to significant improvements in hydrogen pickup and corrosion resistance, the mechanisms by which such alloy improvements occur are still not well understood. In an effort to understand such mechanisms, we conducted a systematic study of the alloy effect on hydrogen pickup, using advanced characterization techniques to rationalize precise measurements of hydrogen pickup. The hydrogen pickup fraction was accurately measured for a specially designed set of commercial and model alloys to investigate the effects of alloying elements, microstructure, and corrosion kinetics on hydrogen uptake. Two different techniques for measuring hydrogen concentrations were used: a destructive technique, vacuum hot extraction, and a non-destructive one, cold neutron prompt gamma activation analysis. The results indicate that hydrogen pickup varies not only from alloy to alloy, but also during the corrosion process for a given alloy. These variations result from the process of charge balance during the corrosion reaction, such that the pickup of hydrogen decreases when the rate of electron transport or oxide electronic conductivity (σe−ox) through the protective oxide increases. According to this hypothesis, alloying elements (either in solid solution or in precipitates) would affect the hydrogen pickup fraction by modifying σe−ox. Because the mechanism whereby these alloying elements are incorporated into the oxide layer is critical to changing electron conductivity, the evolution of the oxidation state of two common alloying elements, Fe and Nb, when incorporated into the growing oxide layers of two commercial zirconium alloys (Zircaloy-4 and ZIRLO) and model alloys (Zr-0.4Fe-0.2Cr and Zr-2.5Nb) was investigated using x-ray absorption near-edge spectroscopy with microbeam synchrotron radiation on cross-sectional oxide samples. The results show that the oxidation of both Fe and Nb is delayed in the oxide layer relative to that of Zr, and that this oxidation delay is related to the variations of the instantaneous hydrogen pickup fraction with exposure time.

Zirconium hydride platelet reorientation in fuel cladding during dry storage and transportation of spent nuclear fuel is an important technological issue. Using an in situ x-ray synchrotron diffraction technique, the detailed kinetics of hydride precipitation and reorientation can be directly determined while the specimen is under stress and at temperature. Hydrided Zircaloy-4 dogbone sheet samples were submitted to various thermo-mechanical schedules, while x-ray diffraction data was continuously recorded. Post-test metallography showed that nearly full hydride reorientation was achieved when the applied stress was above 210 MPa. In general, repeated thermal cycling above the terminal solid solubility temperature increased both the reoriented hydride fraction and the connectivity of the reoriented hydrides. The dissolution and precipitation temperatures were determined directly from the hydride diffraction signal. The diffraction signature of reoriented hydrides is different than that of in-plane hydrides. During cooling under stress, the precipitation of reoriented hydrides occurs at lower temperatures than the precipitation of in-plane hydrides, suggesting that applied stress suppresses the precipitation of in-plane hydrides. The analysis of the elastic strains determined by the shift in position of hydride and zirconium diffraction peaks allowed following of the early stages of hydride precipitation. Hydride particles were observed to start to nucleate with highly compressive strain. These compressive strains quickly relax to smaller compressive strains within 30°C of the onset of precipitation. After about half of the overall hydride volume fraction is precipitated, hydride strains follow the thermal contraction of the zirconium matrix. In the case of hydrides precipitating under stress, the strains in the hydrides are different in direction and trend. Analyses performed on the broadening of hydride diffraction peaks yielded information on the distribution of strains in hydride population during precipitation and cooldown. These results are discussed in light of existing models and experiments on hydride reorientation.

The susceptibility of fuel cladding to failure in the case of a postulated reactivity-initiated accident may be determined by crack initiation within a hydride blister or rim and subsequent crack growth through the thickness of the cladding. This study has determined the fracture toughness of hydrided cold-worked stress relieved Zircaloy-4 sheet subject to through-thickness crack growth at both 25 and 300°C. The experimental approach utilizes a novel procedure in which a narrow linear strip of brittle hydride blister across the specimen width creates a well-defined precrack upon initial loading. The subsequent crack growth resistance is then characterized by four-point bending of the specimen and an elastic-plastic fracture mechanics analysis. At room temperature, the through-thickness fracture toughness (Kq) is sensitive to the orientation of the hydride platelets and Kq≅25 MPa√m for crack growth through a mixed in-plane/out-of-plane hydride field. In contrast, Kq is much nigher (≅75 MPa√m) when the hydride platelets are oriented predominantly in the plane of the sheet and therefore normal to both the crack plane and the crack growth direction. At 300°C, the material exhibits greater ductility as the hydride particles within the matrix resist fracture such that Kq≅83 MPa√m, despite the much lower flow stress of the material.

The intermediate voltage electron microscope‐tandem user facility in the Electron Microscopy Center at Argonne National Laboratory is described. The primary purpose of this facility is electron microscopy with in situ ion irradiation at controlled sample temperatures. To illustrate its capabilities and advantages a few results of two outside user projects are presented. The motion of dislocation loops formed during ion irradiation is illustrated in video data that reveals a striking reduction of motion in Fe‐8%Cr over that in pure Fe. The development of extended defect structure is then shown to depend on this motion and the influence of nearby surfaces in the transmission electron microscopy thin samples. In a second project, the damage microstructure is followed to high dose (200 dpa) in an oxide dispersion strengthened ferritic alloy at 500°C, and found to be qualitatively similar to that observed in the same alloy neutron irradiated at 420°C. Microsc. Res. Tech., 2009. © 2009 Wiley‐Liss, Inc.

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.

Extended abstract of a paper presented at Microscopy and Microanalysis 2009 in Richmond, Virginia, USA, July 26 – July 30, 2009

The experimental study of grain growth in nanocrystalline metallic foils under ion irradiation showed the existence of a low-temperature regime (below about 0.15–0.22Tm), where grain growth is independent of the irradiation temperature, and a thermally assisted regime where grain growth is enhanced with increasing irradiation temperature. A model is proposed to describe grain growth under irradiation in the temperature-independent regime, based on the direct impact of the thermal spikes on grain boundaries. In the model, grain-boundary migration occurs by atomic jumps, within the thermal spikes, biased by the local grain-boundary curvature driving. The jumps in the spike are calculated based on Vineyard’s analysis of thermal spikes and activated processes using a spherical geometry for the spike. The model incorporates cascade structure features such as subcascade formation, and the probability of subcascades occurring at grain boundaries. This results in a power law expression relating the average grain size with the ion dose with an exponent equal to 3, in agreement with the experimental observations. The model is applied to grain growth observed in situ in a transmission electron microscope in a wide range of doses, temperature, and irradiation conditions for four different pure metals, and shown to predict well the results in all applicable cases. Some discussions are also presented on the expansion of the model to the thermally assisted regime. The paper is organized in six sections. Section I gives background and literature review, while Secs. II and III review experimental methods and results for in situ grain growth under irradiation. Section IV derives the model proposed to find the grain-growth equation in the nonthermal regime, and in Sec. V the model is applied to the results. In Sec. VI grain growth in the thermally assisted regime is discussed and Sec. VII presents the conclusions.

Several zirconium alloys with differing weight percentages of Cr, Fe, Cu, and Mo were exposed to flowing, pure supercritical water at 500°C for up to 150 days in an effort to determine their corrosion behavior for consideration in the supercritical water reactor. The weight gains of the alloys were measured, and oxides were characterized after various times. The test results showed a wide range of corrosion behavior depending on the alloy composition and process temperature. The alloys most resistant to corrosion were those containing Cr and Fe, three of which showed protective stable oxides, low corrosion rates, and no breakaway behavior. The ZrCr, ZrCu, ZrMo, and ZrCuMo alloys all exhibited high corrosion rates and non-protective oxides. Analysis of the oxide layer showed that the oxide consisted mostly of monoclinic zirconia (ZrO2). The structure of the oxide-metal interface in the five protective alloys exhibited characteristics that were also seen in protective oxides formed at low temperature, especially the presence of a suboxide layer and an intense (002)T peak at the interface, indicating the presence of a highly oriented tetragonal phase associated with the protective oxide. The change in corrosion kinetics from cubic to linear was directly linked to the size and density of cracks in the oxides.

The fracture behavior of unirradiated Zircaloy-4 containing either solid hydride blisters or hydrided rims has been examined for the contrasting conditions of equal-biaxial and plane-strain tensile deformation at three temperatures (25°, 300°, and 375°C). Cold-worked and stress-relieved Zircaloy-4 sheet containing hydride blisters shows nearly identical failure strains in equal-biaxial and plane-strain tensile deformation for a wide range of blister or rim depths. In all cases, failure strains decrease rapidly with increasing hydride blister or rim thickness, especially in the ≤100 µm range. Test temperature has a significant effect on ductility with failure strains at 300° and 375°C being much greater than at room temperature. The results indicate that the ductility of material containing hydride rims/blisters greater than ≈ 30–40 µm deep is limited by crack growth, which occurs in a mode I manner at 25°C but in a mixed mode I/II manner at ≥300°C (and at higher failure strain levels).

In situ observations in a transmission electron microscope (TEM) were used to study ion-beam enhancement of second-phase precipitation in Zr-Fe nanocrystalline thin films. The free-standing films were prepared by co-sputter deposition with an Fe content of 1.2 at%. TEM diffraction analysis showed that only the hcp Zr crystal structure was present in the as-deposited films. No second phases were detected, although Rutherford Backscattering Spectroscopy (RBS) confirmed a Fe content beyond the solubility limit of Fe in Zr (of the order of ppm). This means the thin films were Zr solid solutions supersaturated with Fe. Heat treatment in the absence of irradiation was observed to cause precipitation of the Zr₂Fe intermetallic phase, but only above 673 K. The same second-phase precipitation can occur at lower temperatures in the presence of ion irradiation. Samples were irradiated in-situ at the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory with Kr ions to fluences in excess of 10¹⁶ ion/cm², at temperatures ranging from 50 to 573 K. Second phase precipitation was detected by electron diffraction patterns and by dark field imaging comparing regions exposed to the beam with regions protected from the beam by the TEM support grid. Precipitation of Zr₂Fe intermetallic phase was observed at all irradiating temperatures above room temperature. In the bulk, this phase is thermodynamically metastable in the range of temperatures investigated (relative to the orthorhombic Zr₃Fe intermetallic phase). The kinetics of the irradiation-enhanced second-phase precipitation was followed by recording the diffraction patterns at regular intervals. The dose to precipitation was found to decrease with increasing irradiation temperature.

TEM investigations in Zr-Fe and Zr-Fe-Ni, Zr-Ni-Cr and Zr-Cr-Fe systems were performed to assess the second particles present in these alloys, either in quasi equilibrium conditions, or after quenching and annealing experiments.

At equilibrium, Zr3Fe was found to dissolve Ni up to 12 at%. Ab initio computations provided formation enthalpies for several Zr-Fe compounds and showed that the Zr2Fe is metastable at low temperature. All of these data allowed us to extend the thermodynamic database ZIRCOBASE; very good agreement between the thermodynamic computations and the experimental evidences was obtained.

Quenching and annealing experiments showed that Zr2Fe is formed for low quenching rates, while Zr3Fe is formed for higher quenching rates.

Last, corrosion results in various conditions show that the influence of second phase particles volume fraction depends on the corroding atmosphere.

Uniform oxidation by the primary circuit water and associated may soon limit the service of Zr alloy fuel cladding in Light Water reactors. Understanding the differences in corrosion rate between alloys based on the microstructure of the protective oxide may allow us to design better alloying materials for severe duty cycle applications. The use of synchrotron radiation microbeam at APS allows the study of these oxide layers with an unique combination of the wealth of diffraction and fluorescence information and the level of spatial resolution obtained. We will discuss some of our experimental results and the potentials of these techniques in solving engineering problems.

Specimen geometries have been developed to determine the mechanical properties of irradiated Zircaloy cladding subjected to the mechanical conditions and temperatures associated with reactivity-initiated accidents (RIA) and loss-of-coolant accidents (LOCA). Miniature ring-stretch specimens were designed to induce both uniaxial and plane-strain states of stress in the transverse (hoop) direction of the cladding. Also, longitudinal tube specimens were also designed to determine the constitutive properties in the axial direction. Finite-element analysis (FEA) and experimental parameters and results were closely coupled to optimize an accurate determination of the stress-strain response and to induce fracture behavior representative of accident conditions. To determine the constitutive properties, a procedure was utilized to transform measured values of load and displacement to a stress-strain response under complex loading states. Additionally, methods have been developed to measure true plastic strains in the gauge section and the initiation of failure using real-time data analysis software. Strain rates and heating conditions have been selected based on their relevance to the mechanical response and temperatures of the cladding during the accidents.

The extended use of Zircaloy cladding in light water reactors degrades its mechanical properties by a combination of irradiation embrittlement, coolant-side oxidation, hydrogen pickup, and hydride formation. The hydrides are usually concentrated in the form of a dense layer or rim near the cooler outer surface of the cladding. Utilizing plane-strain ringstretch tests to approximate the loading path in a reactivity-initiated accident (RIA) transient, we examined the influence of a hydride rim on the fracture behavior of unirradiated Zircaloy-4 cladding at room temperature and 300°C. Failure is sensitive to hydride-rim thickness such that cladding tubes with a hydride-rim thickness >100 μm (≈700 wppm total hydrogen) exhibit brittle behavior, while those with a thickness <90 μm (≈600 wppm) remain ductile. The mechanism of failure is identified as strain-induced crack initiation within the hydride rim and failure within the uncracked ligament due to either a shear instability or damage-induced fracture. We also report some preliminary results of the uniaxial tensile behavior of low-Sn Zircaloy-4 cladding tubes in a cold-worked, stress-relieved condition in the transverse (hoop) direction at strain rates of 0.001/s and 0.2/s and temperatures of 26 to 400°C.

We have conducted a study of second phase particles and matrix alloying element concentrations in zirconium alloys using synchrotron radiation from the Advanced Photon Source (APS) at Argonne National Laboratory. The high flux of synchrotron radiation delivered at the 2BM beamline, compared to conventional X-ray generators, enables the detection of very small precipitate volume fractions. We detected the standard C14 hep Zr(Cr,Fe)2 precipitates (the stable second phase in Zircaloy-4) in the bulk material at a cumulative annealing parameter as low as 10-20 h, and we followed the kinetics of precipitation and growth as a function of the cumulative annealing parameter (CAP) in the range 10-22 (quench) to 10-16 h. In addition, the unique combination of spatial resolution and elemental sensitivity of the 2ID-D/E microbeam line at the Advanced Photon Source at Argonne (APS) allows study of the alloying element concentrations at ppm levels in an area as small as 0.2 X 0.3 μm. We used X-ray fluorescence induced by this sub-micron X-ray beam to determine the concentration of these alloying elements in the matrix as a function of alloy type and thermal history. We discuss these results and the potential of synchrotron radiation-based techniques for studying zirconium alloys.

We present a calculation of the displacement rates and freely migrating defect production caused by neutron and gamma irradiation and their roles on causing irradiation hardening in a BWR core shroud. We find that the neutron displacement rate is much higher than the gamma displacement rate, but that the freely-migrating defects produced by gamma irradiation are significant compared to those produced by neutrons. We evaluate the influence of gamma and neutron irradiation on hardening using a point defect clustering model. We find that the influence of gamma irradiation on radiation hardening in the core shroud is small compared to that for neutron irradiation.

We have conducted a detailed in situ study of phase formation in Zr–Fe metallic multilayers using irradiation and thermal annealing. Metallic multilayers with near equiatomic and Fe-rich overall compositions and with repetition thicknesses ranging from 7.4 to 33 nm were either irradiated with 300 keV Kr ions at various temperatures (from 17 to 623 K) or thermally annealed at 773 K while being observed in situ. The kinetics of multilayer reaction were monitored by following the diffraction patterns. For near equiatomic samples, irradiation causes complete amorphization. The dose to amorphization increases in proportion to the square of the wavelength, indicating a process controlled by atomic transport. Amorphization was also achieved by 900 keV electron irradiation at 25 K showing that displacement cascades are not required. The critical dose to amorphization was independent of temperature below room temperature and decreased above room temperature. The activation energy for this second process is 0.17 eV. For the temperature range studied, diffraction from Zr disappears first, indicating that amorphization takes place in the Zr layer by atomic transport of Fe from the Fe layers. These results are consistent with a combination of simple ballistic mixing at low temperature and either simple diffusion or radiation-enhanced diffusion at higher temperatures. Thermal annealing of the equiatomic samples at 773 K produced the same reaction products with slower kinetics. Ion irradiation of Fe-rich samples did not cause complete amorphization and intermetallic compounds Zr3Fe and ZrFe2 were observed in longer wavelength samples. Amorphization of Fe-rich samples was more sluggish, likely because there was competition with formation of other phases. The reaction kinetics were not proportional to square of wavelength for Fe-rich samples, indicating a process that depends on more than atomic transport. Thermal annealing at 773 K of a long wavelength, 57% Fe sample resulted in intermetallic compounds Zr3Fe and ZrFe2 which amorphized during subsequent irradiation. The ease of amorphization of equiatomic samples relative to Fe-rich samples can be explained by a narrower, single minimum free energy curve for the amorphous phase.

This review provides a comprehensive evaluation of the state-of-knowledge of radiation effects in crystalline ceramics that may be used for the immobilization of high-level nuclear waste and plutonium. The current understanding of radiation damage processes, defect generation, microstructure development, theoretical methods, and experimental methods are reviewed. Fundamental scientific and technological issues that offer opportunities for research are identified. The most important issue is the need for an understanding of the radiation-induced structural changes at the atomic, microscopic, and macroscopic levels, and the effect of these changes on the release rates of radionuclides during corrosion.

The influence of electron energy on the amorphization of ZrCr2 at 25 K was measured. Amorphization was observed at electron energies from 900 to 330 keV. The dose-to-amorphization increases with decreasing electron energy with two steps, one at 700 keV corresponding to the decrease in the Zr displacement cross section due to the approaching displacement threshold of Zr and one at 500 keV corresponding to an appreciable decrease in the displacement cross section of Cr due to the approaching displacement threshold for Cr. At lower electron energies, it is believed that amorphization occurs principally through a secondary displacement mechanism, where light impurity atoms (mainly O) are displaced by electrons and displace in turn the heavier atoms. By fitting the results using electron displacement cross sections, we find the displacement energies in each sublattice to be EdZr=22 eV, EdCr=23 eV, EdO=4 eV.

The High Voltage Electron Microscope (HVEM)/Tandem facility at Argonne National Laboratory has been used to conduct detailed studies of the phase stability and microstructural evolution in zirconium alloys and compounds under ion and electron irradiation. Detailed kinetic studies of the crystalline-to-amorphous transformation of the intermetallic compounds Zr3(Fe1-x,Nix), Zr(Fe1-x,Crx)2, Zr3Fe, and Zr1.5Nb1.5Fe, both as second phase precipitates and in bulk form, have been performed using the in situ capabilities of the Argonne facility under a variety of irradiation conditions (temperature, dose rate). Results include a verification of a dose rate effect on amorphization and the influence of material variables (stoichiometry x, presence of stacking faults, crystal structure) on the critical temperature and on the critical dose for amorphization.

Studies were also conducted of the microstructural evolution under irradiation of specially tailored binary and ternary model alloys. The stability of the ω-phase in Zr-20%Nb under electron and Ar ion irradiation was investigated as well as the β-phase precipitation in Zr-2.5%Nb under Ar ion irradiation. The ensemble of these results is discussed in terms of theoretical models of amorphization and of irradiation-altered solubility.