This paper describes a series of experiments conducted to investigate the in-situ irradiation creep behavior of Zircaloy-4 using proton irradiations as a surrogate to neutrons. The first series of experiments investigated the impact of the initial irradiation-induced defect evolution during the 0 - 0.1 dpa regime on the subsequent in-situ steady-state behavior derived from experimentation. These experiments were conducted at a constant applied load of 85 MPa, a constant damage rate of 1.58 x 10-6-6 dpa/s, and were repeated at both 250 degrees C and 350 degrees C. We also examined the use of a 'conditioning-irradiation' step prior to the creep tests on the results derived from subsequent in-situ proton irradiation creep experiments. By extension, we aimed to further develop and refine optimum testing procedures when using proton irradiations to investigate the in-situ creep behavior of nuclear materials. The second series of in-situ proton irradiation experiments were conducted on two Recrystallized (RX) Zircaloy-4 samples in order to investigate the temperature and stress dependence of irradiation creep. One sample was held at a constant load while the temperature was varied in the range of 275 - 350 degrees C, and the other was held at a constant temperature while stress was varied in the range of 65 - 105 MPa. The associated strains and creep rates were measured at each interval and used to determine an activation temperature of 5000 +/- 1700 K and a stress sensitivity exponent of 4.3 +/- 0.8 for RX Zr-4 over the given temperature and stress ranges. A discussion of potential deformation mechanisms based on competition between bulk diffusion through the lattice and point defect diffusion enabling dislocation climb and glide is given: the relatively high stress dependence suggested the latter is more likely however further investigations will be required to improve the mechanistic understanding. The results presented in this manuscript, including activation temperature, stress sensitivity, and associated creep rates, determined through proton irradiation investigations are closely comparable to those determined from neutron irradiation experiments found in the literature for similar zirconium alloys at similar temperatures and stresses. Coupled with the primary-secondary creep behavior investigations presented, this demonstrates the usefulness of this approach to estimate neutron-irradiation equivalent creep behavior using in- situ proton irradiation.

The effect of prior proton irradiation on the subsequent corrosion and dealloying of Alloy 800 was investigated in a 50 wt. % NaOH caustic solution at 140 degrees C. Dealloying of irradiated (1 dpa) and non-irradiated Alloy 800 was characterized at the nanoscale to provide mechanistic insight. A porous surface film enriched in Ni was formed on both irradiated and non-irradiated materials due to selective dissolution of Fe and Cr. However, the microstructure and chemistry in the dealloyed layer was altered by irradiation, including deeper dealloyed layer penetration, finer porosity, and changes in the classical core-shell ligament structure in the irradiated material.

The oxidation of Ni-Cr-Al model alloys was investigated after exposure to a 480 degrees C hydrogenated steam environment. Low angle grain boundaries (LAGBs) and the studied coincidence site lattice grain boundaries (CSLBs) showed intergranular oxidation, while high angle grain boundaries (HAGBs) were protected by an external oxide. The intergranular oxide consisted of a central Al-rich oxide, surrounded by a Cr-rich oxide. Al contribution in intergranular and external oxidation was enhanced in Ni-Cr-Al alloys compared to Ni-Al alloys. Thus, the third element effect is applicable at low homologous temperature along short-circuit diffusion pathways, in this case grain boundaries.

Irradiation induced growth is a constant volume shape change that occurs without externally applied stress that is observed in some materials under irradiation damage. Proton irradiation was carried out on a pure Zr sample to many dpa, to enable investigation of microscale aspects of the irradiation growth phenomenon. Irradiation induced a significant surface morphology change in the irradiated area, which is believed to be the result of the anisotropic growth behavior of the hcp structured Zr. The localized strain that developed was characterized by Electron back scatter diffraction (EBSD), on both the irradiated surface and on a cross sectional through -thickness plane. It was found that there is a correlation between the amount of local deformation and level of misorientation existing between two adjacent grains. The irradiation induced defect microstructure was characterized by transmission electron microscopy (TEM), showing (a) and (c) component loops similar to that generated by neutron irradiation in literature. Lastly, site specific focused ion beam (FIB) TEM lift -outs were prepared on local grain boundaries to investigate the origin of the localised deformation.

A material of magnesium containing 5 wt % carbon is brought to 1000 hydrogen sorption cycles with systematic sampling for material characterization. The Mg-5C material is cycled isothermally at 330 degrees C using an in-house hydriding rig. Samples removed from the reactor are analyzed using powder X-ray diffraction (XRD) and scanning electron microscopy (SEM) with additional transmission electron microscopy (TEM) on a subset of samples. These data are compared to cycling performance with the aim of finding correlations between material performance and structural evolution. The microstructure is shown to evolve in two stages. The first is the rapid transformation of the as-milled material into a high-aspect ratio structure. The second stage is the slow agglomeration and densification of the evolved structure via sintering. After 1000 cycles, the material retains nearly 88% of its peak hydrogen capacity.

This work investigates the potential metastable versus stable nature of the γ-hydride in Zircaloy-2. Specimens were hydrided, and hydrides were precipitated through water-quenching. Synchrotron X-ray diffraction of the water-quenched sample revealed a diffraction peak with the d-spacing value of ∼2.70 Å. While this peak is conventionally attributed to {111}-γ planes, it could also stem from the (0004)-ζ plane. To clarify this ambiguity, the crystal structure of nano-hydrides was characterized by nano-beam electron diffraction (NBED) and electron energy-loss spectroscopy (EELS). While EELS detected nano-hydrides with plasmon energy (PE) values associated with the ζ-,γ-, and δ-phases, suggesting all three types of phases might be present, complementary NBED analysis revealed that regardless of the measured PE values, the examined nano-hydrides were of only γ- or δ-nature.Repeating the heating/quenching cycles reduced the γ-phase volume fraction until its complete disappearance after three cycles. δ-phase, however, was observed after each heating/quenching cycle. This observation, in accordance with previous reports, indicates the γ-phase is metastable in Zircaloy-2, such that even during water-quenching (which is conventionally believed to facilitate the formation of γ-phase) only δ-hydrides form in cases where microstructural conditions are suitable.Diffraction line profile analysis and transmission electron microscopy revealed an increase in dislocation density during the first heating/quenching cycle, with no noticeable variations during subsequent cycles. A mechanism is proposed that links microstructure (i.e., dislocation structure) evolution during heating/quenching cycles to the suppression of the (metastable) γ-phase.

Alloy X-750 was exposed to a 480 degrees C CO-CO2 reducing gas mixture at an oxygen partial pressure below the NiO dissociation pressure. Post-exposure analytical transmission electron microscopy (ATEM) analysis confirmed the occurrence of preferential intergranular oxidation (PIO) and internal oxidation which resulted in the formation of Ni nodules. Thus, reducing CO-CO2 mixtures can promote PIO in Alloy X-750, similar to Alloy 600 exposed to similar oxygen partial pressure but in H-2-steam environment. In Alloy X-750, intergranular oxidation of Ti and Al were observed to occur prior to Cr oxidation, while the presence of intergranular Cr carbides hinders PIO penetration depth.

The zirconium (Zr) zeta-hydride phase was first proposed in 2008. While electron diffraction experiments characterized a trigonal crystal structure for this phase, formation-energy calculations have predicted that trigonal is not a stable structure for that. In this work, we attempt to tackle this discrepancy between experimental and computational studies. For this purpose, nano-beam electron diffraction (NBED) and electron energy-loss spectroscopy (EELS) were utilized to characterize the crystal structure of nano -hydrides in annealed Zircaloy-2. Diffraction patterns (DPs) collected from multiple axes of precipitates (and identical to those reported for the zeta-hydride in previous works) are shown to be composed of typical alpha-Zr and 8 -hydride reflections along with additional reflections that did not belong to either phase. It is shown that precipitates were 8 -hydride and extra reflections did not originate from a trigonal structure. Instead, it is proposed that they originated from either 8 -hydride and a thin surface (probably Zr-oxide) phase or double-diffraction scattering between the alpha-Zr and 8 -hydride. Previous works reported a plasmon energy (PE) value of 17.4 +/- 0.1 eV for the zeta-phase. PE maps in this work confirmed the 8 -nature of the examined nano-hydrides and revealed that the measured 17.4 eV PE value belonged to interfacial ribbons as a normal interface effect, instead of a zeta-phase. Hence, as predicted by computations, it is concluded that the trigonal structure appears to not be the appropriate crystal structure for nano-hydrides and reported observations of the zeta-hydride in the past could, in fact, have been 8 -phase that was misidentified.

Inconel & REG; X-750 has been implanted with helium at a temperature of 400 degrees C to obtain an average concentration of [He] of about 3000 atomic parts per million (appm). The implantation resulted in the formation of a low density of point defect clusters in the form of dislocation loops and a high density of He-stabilized cavities. Uni-axial tensile tests were conducted at room temperature on both the helium-implanted and non-helium-implanted materials. There was a reduction in ductility and an increase in yield strength after helium implantation. The fracture surface of the helium-implanted-material exhibited a mixed failure mode of inter-granular and transgranular fracture. TEM observations showed that the trans-granular fracture is the result of failure along twin/martensite platelets that were present in the material prior to implantation. The interfaces of these platelets with the austenite matrix, which are bordered either by incoherent twinned material or thin layers of & epsilon;-martensite, are paths for crack propagation that may be enhanced by He segregation at the interfaces. The results are discussed in terms of the implications for observations of channel fracture in neutron irradiated austenitic stainless steels and Ni-alloys.

The intergranular oxidation of binary Ni-xAl (x = 4, 7 at.%) and Ni-yCr (y = 11, 15 at.%) model alloys was studied in a 480 degrees C hydrogenated steam environment. It was observed that the intergranular oxidation sus-ceptibility in Ni-Al and Ni-Cr alloys is different. In Ni-Al alloys, HAGBs are very susceptible to intergranular oxidation, while low energy grain boundaries behave similar to bulk material. In Ni-Cr alloys, LAGBs and CSLBs are susceptible to intergranular oxidation, while HAGBs are protected by an external oxide. The structure of the intergranular oxide was also different between Ni-Cr and Ni-Al alloys; the intergranular oxide formed in Ni-Al alloys was sharper and penetrated to greater depth compared with Ni-Cr alloys. Comparisons are made to previous testing of alloy 600 (Ni-16Cr-Fe) in similar environments, highlighting the potential effect of residual Al and Ti on the intergranular oxidation of the alloy.

The crystal structure of the interfacial areas of a & delta;-FCC water-quenched nano-hydride was characterized by electron energy-loss spectroscopy (EELS) and electron diffraction. EELS revealed ribbons with plasmon energy (PE) values of 17.4 & PLUSMN; 0.01 eV and 18.3 & PLUSMN; 0.01 eV (nominally characteristic of the & zeta;- and & gamma;-hydride phases, respectively) in the interfacial area between the & delta;-core and & alpha;-Zr matrix. Electron diffraction patterns (DPs) ob-tained from the <2110> axes of the interface contained reflections that could be indexed as {0001} reflections of the & zeta;-phase. Such & zeta;-type reflections, however, disappeared after tilting the interface away from the <2110>-axes; implying that they originated from sources other than a hypothetical & zeta;-phase. Moreover, electron DPs obtained from multiple zone axes of the interface, did not show characteristic {110}/{112} superlattice reflections of the & gamma;-phase. These results ruled out the existence of & zeta;- and & gamma;-phases in the interface (down to the spatial resolution of the utilized techniques) despite the measured plasmon energy values. Subsequently, dielectric theory was utilized to clarify the origin of interfacial ribbons with PE values char-acteristic of the & zeta;- and & gamma;-phases. Dielectric functions of the & alpha;-Zr and & delta;-core were extracted from the energy-loss spectra of the two phases, to simulate the energy-loss functions across the interface. Simulations suggested that the observed interfacial ribbons in EELS maps (with PE values of 17.4 & PLUSMN; 0.01 eV and 18.3 & PLUSMN; 0.01 eV) could have stemmed from the delocalized nature of plasmon vibration and the effect of interface on shifting the plasmon vibration frequency, but not necessarily from the existence of the & zeta;- and & gamma;-phases.

The mechanical properties of delta-zirconium hydrides were studied using a combination of nanoindentation techniques and numerical and analytical models. Two different alloys, containing different forms of delta-zirconium hydride (blister and rim), were analyzed at room temperature and at elevated temperatures up to 300 C. A reduction of hardness from RT to 300 C showed an approximately linear relationship with the change in temperature. For the delta-zirconium hydride rim, the hardness reduced from 3.73GPa at RT to 1.49GPa at 300 C, a reduction of 61 % . For the delta-zirconium hydride blister, hardness decreased from 3.5GPa at RT to 1.91GPa at 300 C, a reduction of 45 % . The yield stresses were calculated using two approaches: a numerical model proposed by Johnson to calculate values at RT, and another proposed by Dao for values up to 300 C. At RT, the yield stress for the delta-zirconium hydride rim obtained was 883 +/- 21MPa, 224MPa at 300 C. The delta-zirconium hydride blister showed a reduction from 879.5 +/- 10.5MPa at RT to 324MPa at 300 C, a reduction in 61 % and 45 % for the hydride rim and blister, respectively. Through this combination of experimental and modeling techniques, this study showed the variation of hardness and yield stress in complex systems, and in different environments, using nanoindentation as the main characterization technique.(C) 2021 Elsevier B.V. All rights reserved.

The effect of manufacturing routes (heat treatment done by a conventional β-quenching (BQ), or slow cooling (SC) prior to hot extrusion), and location along the tube (front end (FE) and back end (BE) emerging from the press) on hydride precipitation behavior in Zr-2.5Nb micro pressure tubes were evaluated. After being hydrided to 100 wppm H, stepped micro pressure tube samples with three kinds of microstructure (BQ-BE, SC-FE, SC-BE) were pressurized to obtain nominal hoop stresses ranging from 65 MPa to 350 MPa. All the samples were heated to 350°C to dissolve all of the hydrogen, followed by cooling under stress to allow hydride reorientation to occur. Almost complete reorientation was observed in the BQ-BE sample at stress levels of above 100MPa, while the SC-FE and BE samples showed stronger resistance to reorientation. After subsequent reheating to 350°C for 1 hour or 24 hours and cooling without stress, little memory effect was found for the re-orientated hydrides in the BQ-BE sample. However, a clear memory effect was observed in the SC-FE and SC-BE sample in the range of angles where hydrides were naturally present.

A new design for an ion beam energy degrader has been developed and tested at the Reactor Materials Testing Laboratory at Queen's University. With the use of a dual rack multi-foil system, it is now possible to implant helium uniformly throughout a nickel target to a depth of 33 mu m with helium ions. Aluminium foils are used in the degrader with thicknesses ranging from 2-84 mu m, causing deceleration of the inci-dent particle and thus a shallower penetration depth in the target with increased Al foil thickness. The design was successfully tested with the helium implantation of nickel superalloy, Inconel X-750. Using transmission electron microscopy, helium bubbles were imaged following an implantation of 80 0 0 appm helium. A continuous formation of helium bubbles was observed with implantation depth, matching the predicted implantation range. The prediction for helium implantation using the aluminium foils to de-grade the incident particle is carried out with data generated from SRIM calculations and an inhouse optimization code.(c) 2022 Elsevier B.V. All rights reserved.

The behavior of radial hydrides in Zr-2.5Nb pressure tube samples at elevated temperature has been studied. The aim of this work is to investigate the stability of stress-reoriented hydrides under heating without an external applied load. Using in-situ heating at beamline, 1-ID at Argonne National Laboratory, USA, it was shown that two populations of hydrides existed in the sample, naturally oriented and reoriented hydrides. At a relatively low temperature of 180 degrees C it was observed that the reoriented hydrides completely dissolved, leaving only the naturally oriented hydrides in the sample, when heated without an applied load. On subsequent cooling of the sample, only naturally oriented hydrides were observed. These observations were confirmed using in-situ scanning electron microscopy heating experiments and post-mortem transmission electron microscopy studies. The results presented here offer a potential route for ductility recovery of pressure tube materials where radial hydrides have been shown to significantly reduce fracture toughness of components. (c) 2022 Elsevier B.V. All rights reserved.

In the binary Zr-H, Ti-H systems, there are three common hydride phases, namely gamma-, delta-, and epsilon-hydride, with different ratios of H to Zr/Ti atoms. Among them, the relative stability of gamma-hydride remains controversial, although it has been observed in many studies. In this study, we demonstrate that gamma-hydride is the only stable hydride phase in high purity Zr and commercial purity Ti, using electrolytically loaded hydrogen and ex-situ synchrotron x-ray diffraction. Further, in-situ synchrotron heating/cooling experiments were conducted with hydrided pure Zr sample demonstrating the co-existence of gamma- and delta-hydride phase, from a furnace cooling heat treatment followed by three days room temperature aging. It is shown that during heating gamma-hydride dissolved first, while delta-hydride remained stable at temperatures up to 280 degrees C, followed by dissolution of delta-hydride until all hydrogen was in solution. Under subsequent slow cooling, delta-hydride precipitated first, followed by a slow formation of gamma-hydride when temperatures were below 210 degrees C, which continued during a subsequent 60 hours aging at room temperature. The phase stability when hydride forms as precipitates in Zr is therefore determined to be delta-hydride as the high temperature phase, and gamma-hydride as the low temperature phase. The microstructural characteristics of gamma-hydride in Zr and Ti were studied by electron microscopy. The observed co-existence of gamma- and delta-hydride as well as change from delta to gamma during long term room temperature aging in the TEM confirmed the synchrotron data. (C) 2020 Elsevier B.V. All rights reserved.

A failure analysis was performed on an Alloy C-276 pull rod which underwent unexpected brittle, intergranular fracture after exposure to 280°C to 300°C aqueous solutions designed to replicate secondary side environments in nuclear energy systems: Pb-containing alkaline (pH300°C 8.5 to 9.5), and sulfate-containing acidic solutions (pH280°C 3 to 5). The component was characterized using advanced electron microscopy methods to demonstrate the benefits of these techniques for determining the nanoscale chemical, mechanical, and material factors contributing to failure, and to provide insight into the mechanisms of stress corrosion cracking (SCC) responsible for failure. Site-specific transmission electron microscopy specimens containing crack tips were prepared using focused ion beam. Nanoscale chemical characterization methods revealed that Pb was present in some oxidized regions of cracks, suggesting that the element may be inhibiting or impairing the passivity of the Cr-rich oxide. Complementary nanoscale microstructural analysis was performed. At an intergranular to transgranular cracking mode transition, it was observed that the transgranular crack (and corrosion process) propagated along the (110) crystallographic plane. Also, the cracking mode was highly dependent on the tensile stress direction relative to grain boundary orientation, the crystallographic orientation of grains, and geometrically necessary dislocation structures. A comparison of results with proposed mechanisms for SCC of Ni alloys in similar environments are discussed; the highly directional nature of cracking is consistent with a slot-tunnel corrosion mechanism.

The regions of CANDU® pressure tubes near the rolled joint (RJ) are the un-supported portions of the inlet and outlet ends just inboard of where the tubes make contact with the 403 stainless steel end fitting sections. Compared to the main body of the pressure tube, in-service conditions are more complex with large gradients in the flux, significant hydrogen isotope concentration profiles and varying operating temperatures. In order to increase the amount of data related to microstructure evolution, the microstructures near rolled-joint regions from an ex-service CANDU pressure tube were investigated. The results show that irradiation-induced effects on the microstructure are more significant than the temperature influence near the inlet rolled-joint region (that operates at lower temperatures). Thermal effects play a more dominant role than the flux near the outlet region (that operates at higher temperatures). Zirconium hydrides aligned in both circumferential and radial directions of the tube have been observed. The zirconium hydrides have an orientation relationship of (0002)αZr//(1¯1¯1)ZrH, and [21¯1¯0]αZr//[011]ZrH with the α-Zr matrix; the hydride phases were identified as gamma (γ) hydride by observing superlattice reflections on the (110) planes in the electron diffraction patterns. The presence of γ hydrides at room temperature in the studied tube may have technical importance because in most of the previous research, δ hydrides were considered as the stable phase that would precipitate during cooling at reactor relevant rates. This assumption has been applied to both delayed hydride cracking (DHC) and fracture toughness experiments and during development of related models. It is critical to elucidate the phases of zirconium hydrides at relevant in-service conditions to enable representative experimental testing and modelling that can be used to verify and/or improve the predictive capability of material evolution and cracking behavior of zirconium alloys in nuclear reactors.

Environmentally assisted cracking (EAC) initiation tests were carried out by subjecting Alloy 800 tensile specimens to 0.55 mol/kg SO42- solution, pH(280 degrees C) 3, at 280 degrees C using slow rise-time cyclic loading and in-situ crack detection. EAC, intergranular corrosion (IGC), and pitting were observed. Transmission electron microscopy analysis revealed Ti- and Cr-rich oxides in cracks, and sulfur incorporated in oxide(s) or as sulfide compound(s). This oxide/sulfide film is likely impaired, producing a slip dissolution-type EAC mechanism. For pitting, only a nano-scale sulfur layer was identified at pit-metal interfaces. This high surface coverage of adsorbed sulfur limited oxide nucleation and accelerated metal dissolution.

Alloy 800 was exposed to a 330 degrees C Pb-containing, mildly caustic (pH(330 degrees C) 9.5) environment producing SCC. High resolution analytical TEM characterization of crack tips provided evidence to elucidate the Pb-caustic SCC mechanism and highlight a continuum of classical caustic SCC, facilitated at lower pH by the presence of Pb. Successive regions of Ni-enrichment and Cr-rich oxides were observed along the crack with selective Fe dissolution ahead of crack tips, de-alloying. Also, Pb is observed at oxide-metal interfaces. A hybrid film-rupture/de-alloying SCC mechanism is proposed, with Pb at oxide-metal interfaces acting to impair the passivity of the metastable Cr-rich oxide.