Irradiated Microstructure Research Articles

Effect of irradiation on the corrosion of 304 stainless steel in pressurized water reactor (PWR) simulated water chemistry

In this study, the corrosion behavior of pre-ion-irradiated 304 SS is compared to that of un-irradiated 304 SS in simulated Pressurized Water Reactor (PWR) conditions for both early stages of corrosion and longer exposure times. Transmission electron microscopy is used to characterize the oxide scale. The study shows that although the pre-irradiated microstructure does not change the nature of the oxide and its dual-layer structure (i.e. inner/outer oxide layers) formed on 304 SS, the path of growth of the inner oxide layer is affected by the irradiated microstructure resulting in oxide protrusions into the matrix; a mechanism is proposed.

The effect of hydrogen co-injection on the microstructure of triple ion irradiated F82H

The reduced activation ferritic/martensitic steel F82H-IEA was studied through a systematic set of dual (Fe2++He2+) and triple (Fe2++He2++H+) ion irradiations to determine the effect of hydrogen co-injection with helium and radiation damage on the irradiated microstructure and in particular, cavity evolution. Irradiations were conducted in four distinct permutations from a set of nominal irradiation parameters of 50 dpa, 1 × 10–3 dpa/s, 500 °C, and 40/10 appm/dpa of H/He or 10 appm/dpa of He. A temperature series of dual and triple ions ranging from 400 to 600 °C was conducted where swelling values of 1.3–2.4x were obtained with the co-injection of hydrogen. A damage level (50 to 150 dpa) and damage rate (2.5 × 10–4 to 2.5 × 10–3 dpa/s) series were also conducted where hydrogen was again observed to cause a ∼1.2–1.8x increase in swelling in triple ion irradiation versus dual ions. The effect of hydrogen was also observed to increase the overall steady state swelling rate from 0.02%/dpa in dual ions to 0.034%/dpa under triple ions at 500 °C by 150 dpa. A final series of triple ion experiments was conducted at 450 °C and 500 °C at 80 appm/dpa of H where swelling was suppressed by 50% compared to the dual ion irradiated case due to over-nucleation of small bubbles. The effect of hydrogen on bubble nucleation was to stabilize higher densities of smaller helium bubbles possibly through increasing nucleation rates and reduction of the stable bubble radius. The effect on swelling was in the promotion of higher densities of larger voids experiencing bias-driven growth. Hydrogen was shown to have a moderate influence on the nucleation and growth of dislocation loops, although these effects appear to be convoluted by the changes in cavity microstructure dominating sink strength across conditions. Hydrogen co-injection was also loosely correlated with enhanced dissolution of M23C6 precipitates across temperature and damage level regimes.

Effect of H+ pre-irradiation on irradiation hardening and microstructure evolution of FeCoCrNiAl0.3 alloy with He2+ irradiation

FeCoCrNiAlx-based high entropy alloys (HEAs) exhibit excellent high-temperature strength, good radiation tolerance and corrosion resistance. However, studies on light-ion irradiation damage are relatively lacking. The influence of hydrogen and helium effect on HEAs deserves further study, given their complex interaction during different sequence irradiation at different irradiation conditions. In the present study, FeCoCrNiAl0.3 HEA was irradiated under two different modes at 723 K, one is single He2+ and the other is pre-iradiation of H+ followed by He2+. Microstructure evolution and irradiation hardening were studied by transmission electron microscopy and nanoindentation technique. There is no distinct difference of the irradiation hardening rate between sequential H+-He2+ and single He2+ irradiated samples. A large number of irradiation-induced nano-defects were observed after pre-irradiation of H+, which act as sinks to trap the subsequent helium atoms in the H+-He2+ irradiated sample. Fine helium bubbles and dislocation loops with high density dispersed in both irradiated samples. Compared with single He2+ irradiation, pre-irradiation of H+ resulted in a decrease in bubble density and an increase in loop density in H+-He2+ irradiated sample, but had little effect on their size. It is speculated that the critical concentration of bubble formation originating from H-He-V clusters is supposed to be higher than that of HeV clusters at elevated temperatures in FeCoCrNiAl0.3 alloy, leading to the aggregation and redistribution of He atoms at the H-He-V clusters rather than the formation of stable bubble nucleus after succeeding He2+ irradiation. The influence of pre-irradiation of H+ on FeCoCrNiAl0.3 alloy on the formation of helium bubbles and the relationship between bubble, loop and irradiation hardening are discussed.

Irradiation microstructures and hardening in commercial nuclear reactor pressure vessel steels at high extended life fluence

Nuclear reactor lifetimes up to and beyond 80 years may be limited by nano-scale Cu-Mn-Ni-Si precipitates (CRPs and MNSPs) that form under neutron irradiation in pressure vessel (RPV) steels, resulting in hardening and ductile to brittle transition temperature increases, or embrittlement. Here, Atom Probe Tomography (APT) is used to characterize precipitation in nine actual RPV steels, irradiated in the ATR-2 test reactor, to a high fluence 1.38 × 1020 n/cm2 at 292 °C and 3.68 × 1012 n/cm2 s. APT quantifies the relation between a wide range surveillance alloy Cu, Ni, Mn, and Si contents and the corresponding precipitate volume fractions, number densities, sizes, and compositions. The precipitate volume fraction (fp) is dominated by the dissolved steel Cu and Ni contents. Irradiation hardening shows the expected dependence on the √fp. The APT results also provide an opportunity to evaluate flux effects by direct comparisons with corresponding archival data from low flux surveillance irradiations. Most notably, this work establishes a strong microstructural and mechanistic foundation for the new data-driven Odette, Wells, Almirall, and Yamamoto (OWAY) model to predict high fluence RPV steel embrittlement.

Extending damage accumulation of commercial reactor irradiated 316 stainless steel with ion irradiation

Austenitic 316 stainless steel flux thimble tubes (FTTs) removed from a commercial pressurized water reactor (PWR) were re-irradiated using nickel ions to evolve the irradiated microstructure to higher damage levels than those achieved in reactor. The microstructures of the neutron-plus-ion irradiated samples were compared with those of neutron irradiated only samples at the same doses to evaluate the effectiveness of ion irradiation to extend the damage range and match that created in reactor. Ion irradiations effectively captured, both qualitatively and semi-quantitatively, the evolution of dislocation loops, nanocavities, nanocluster size and density, and radiation-induced segregation (RIS) with dose. Only Ni-Si clusters were observed in ion irradiated FTT samples while Ni-Si-Mn clusters that were frequently observed in reactor irradiated samples above ∼41 dpa were absent in ion irradiated samples probably due to ballistic mixing at the high ion irradiation damage rate. Two samples were ion irradiated to ∼160 dpa to predict the irradiated microstructure that may develop due to an increase of the damage accumulation by extending a PWR life from 40 years to 60 years. The extended ion irradiation results indicated that significant microstructure changes were unlikely to occur even to very high doses near common PWR relevant temperature conditions. Overall, the microstructure resulting from ion irradiation following neutron irradiation matched that of neutron to the same dose reasonably well in most cases.

Cluster dynamics of proton irradiated RPV steel at room temperature

The life of reactor pressure vessels in nuclear power plants will be extended to 60 years, while the irradiation data of materials from commercial and experimental reactors are rare and expensive. There is an urgent need to solve how to accurately predict the changes in microstructure and mechanical properties of high-dose irradiated materials of RPV. The cluster dynamics method can be used to establish the relationship between nanostructure and macro properties and to study its evolution over a long period of years. Based on the average rate field theory, a cluster dynamics model was developed for the evolution of irradiated microstructure of RPV steel, which considers both homogeneous and heterogeneous nucleation of solute clusters. The model was successfully used to simulate the influence of the proton irradiation-induced cluster size and number density of RPV model steel with a low copper content at room temperature.

Defect-specific strength factors and superposition model for predicting strengthening of ion irradiated Fe18Cr alloy

A high-purity binary alloy Fe18Cr was subjected to heavy-ion irradiation in order to provide improved understanding of the irradiation effect on radiation hardening associated with dislocation loop and network formation, α’ precipitation. The specimens were irradiated with 8 MeV Fe ions (∼2 μm ion range) to midrange doses of 0.37 and 3.7 dpa at 300, 350 and 450 °C using dose rates of ∼10−5–10−3 dpa/s. Nanoindentation testing was performed to extract the bulk equivalent hardness at low depth from these ion irradiated specimens. High-quality transmission electron microscopy (TEM) images were acquired by utilizing a flash electropolishing method on the Focused Ion Beam (FIB) prepared liftouts. An accurate calculation of the strength factor was provided based on detailed TEM characterization of the irradiated microstructures. A newly refined hardening superposition method was applied to combine the strengthening components from each microstructure. The good agreement between microstructure predicted strength and measured strength further demonstrated the fidelity to use hardening model to quantify the mechanical properties of ion irradiated materials.

Irradiation creep measurement and microstructural analysis of chromium nitride–coated zirconium alloy using pressurized tubes

Environmental barrier coatings for Zr-based materials are currently under development to reduce oxidation and embrittlement in light-water reactors. Chromium nitride is one such candidate for this application, particularly as accident-tolerant fuel cladding. However, quantifying the impact of coatings on the irradiation-induced creep of zircaloy (Zry) is critical as this mechanism often exceeds thermal creep rates under light-water reactor operating conditions and can be a limiting design characteristic. Additionally, examining irradiation effects in the microstructure at the coating interface is key to understanding the compatibility of the material system. To accelerate the experimental measurement of irradiation creep and microstructure evolution in CrN-Zry, compact, pressurized creep tubes were fabricated and irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Miniature, thin-walled rodlets fabricated from annealed Zr-Sn barstock were coated with CrN using physical vapor deposition (PVD) to nominal thicknesses of 4 and 8 μm. Coated and uncoated rodlet specimens were internally pressurized and welded, generating nominal circumferential hoop stresses of 0, 90, or 180 MPa under 300∘C irradiation conditions. Twelve specimens were measured diametrically prior to irradiation using a low-cost, automated, contactless laser profilometer developed for this work. Specimens were irradiated in sealed capsules for one 25-day HFIR cycle, accumulating approximately 1.8×1021n/cm2 fast fluence (En>1.0MeV). The irradiated samples were retrieved and remeasured using the same profilometry system in a shielded hot cell facility. Irradiation creep between specimens was compared using standard statistical tests and showed that both thicknesses of CrN coating had a negligible effect on the irradiation creep strain of the Zry material. Microstructure characterization of pre- and post-irradiated CrN-Zry specimens showed minimal changes due to irradiation but did show a substantial O-rich region at the Zry-CrN interface.

Analytical flow equation for irradiated low-alloy steels established by multiscale modeling

In this paper, we show how results of multiscale modeling can be condensed to establish a simple analytical flow equation for irradiated low-alloy steels. The flow equation accounts for temperature, strain rate, the initial and irradiation microstructure. Starting from a complete set of constitutive equations describing crystal plasticity of irradiated ferritique steels, simple assumptions and approximations are used to integrate the equations over the homogenized polycrystal. The final flow equation incorporates the lattice friction resistance, the Hall-Petch effect, the forest strengthening and irradiation hardening. Every component is established by multiscale modeling at lower scales. Irradiation hardening is attributed to solute clusters. Their size and density are determined in experiment, their shear resistance computed by atomistic simulations and their contribution to the flow stress assessed using dislocation dynamics simulations. The formation of solute clusters is found to increase the flow stress and the strain hardening modulus. Irradiation hardening decreases with temperature and with the yield stress of the unirradiated material. Predictions of the tensile curves and irradiation hardening are compared with many experimental results. A special section is dedicated to the interpretation of irradiation hardening measured in a surveillance program.

The role of Cr, P, and N solutes on the irradiated microstructure of bcc Fe

The objective of this study is to understand irradiation-induced and assisted defect evolution in binary body center cubic (bcc) Fe-based alloys. The broader class of bcc ferritic alloys are leading candidates for advanced nuclear fission and fusion applications, in part due to their exceptional void swelling resistance. However, their irradiated microstructure evolution is sensitive to solute species present, since these solutes can act as traps for irradiation-induced defects due to the surrounding tensile or compressive stress fields. Here, three alloys (Fe-9.5%Cr, Fe-4.5%P, and Fe-2.3%N) are selected for study because they systematically exhibit varying solute sizes and solute positions (i.e., substitutional or interstitial). Ex situ and in situ ion irradiations reveal that Fe-P has a considerably finer and denser population of irradiation-induced defects than Fe-Cr and Fe-N at the same irradiation conditions, which is attributed to strong defect trapping at undersized substitutional P, consequently hindering the development of extended defects. Meanwhile, oversized substitutional solutes (e.g., Cr) and interstitial solutes (e.g., N) may also suppress dislocation loop development due to weak solute-defect trapping.

Possible impacts of Mo chemical banding and second phase impurities on the irradiation behavior of monolithic U-10Mo fuels

This study investigated the microstructural behavior of both full-size and mini-size monolithic U-10Mo fuel plates irradiated to high burnup with a focus on the evolution of the second phase impurities in monolithic U-Mo using scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), and wavelength dispersive spectroscopy (WDS). Key indicators of possible mechanical and thermal compromise include cracks, large fission gas porosity, and interconnection of fission gas pores. For the fission densities evaluated in this work (3.5 × 1021 fissions/cm3 -5.1 × 1021 fissions/cm3), fine porosity can develop along the UC phase boundary; however, the size of the fission gas pores is no more than those observed in the U-Mo fuel phase. Other inclusions such as Si-rich second-phase impurities found in the as-fabricated microstructure were difficult to resolve post-irradiation because they can become overshadowed by porosity development in the fuel phase. Additionally, the presence of a Fe-rich sublayer formed in the Zr diffusion layer during fabrication remained enriched in the Zr layer in the irradiated U-10Mo microstructures near the U-Mo/Zr interface; however, based on the burnup assessed in this study the identified impurities did not appear to contribute to notable microstructural degradation.

Development of hydrothermal corrosion model and BWR metal coating for CVD SiC in light water reactors

SiC/SiC fiber composites with CVD SiC overcoats are potential candidates for light water reactor advanced accident tolerant cladding materials. Understanding its corrosion kinetics in Light Water Reactor (LWR) conditions is essential to evaluate the concept's viability. Existing models only account for the temperature and oxygen concentration effect on the hydrothermal corrosion behavior applicable to LWR operating conditions. However, the development of a general corrosion rate for CVD SiC that accounts for the impact of irradiated microstructure, flow rate, electrical resistivity, pH, and surface roughness is critical for the practical realization of SiC/SiC-based cladding concepts. After a rigorous experimental campaign, this work updates the existing hydrothermal corrosion model to predict hydrothermal corrosion in LWRs. Numerical radiation and coolant chemistry analysis for LWRs conducted based on the updated corrosion kinetic models suggests that CVD SiC is likely a viable environmental barrier coating for Pressurized Water Reactors while questionable for Boiling Water Reactors (BWR) when the effect of irradiation damage on SiC corrosion is considered. An effective mitigation strategy for the double-layer metal coating is proposed for BWR applications. The double-layer metal coating comprising a FeCrAl overcoat with an intermediate Cr bond coating was observed to provide a stable protective barrier against SiC dissolution in BWR conditions. The proposed metal coating was also fully adherent following quench and burst tests.

Atomic-scale study of He ion irradiation-induced clustering in α-Zirconium

Zircaloy-4 alloy specimens consisting of α-Zr grains were irradiated on a tandem accelerator with 5 MeV He ions to a fluence of 5 × 1021 ions m − 2 at different temperatures. Defect clusters and He bubbles were observed in the TEM micrographs of the irradiated microstructure. The small defect clusters and He bubbles were found to exhibit spherical shapes whereas the large defect clusters showed plate-like shapes. Molecular dynamics simulations and crystallography analyses revealed that the defect clusters are mainly composed of Zr interstitials. During He-ion irradiation, the clustering of Zr dumbbell interstitials results in the formation of small and spherical defect clusters. With further irradiation, the Zr interstitials prefer to occupy the octahedral sites that align along the {12¯14} planes with other site interstitials, forming cluster sections. These initial planar clusters then grow by extending along the <101¯0> directions and by forming more cluster sections along the <12¯16> directions, leading to large plate-like defect clusters on the {1¯21¯1}planes. He bubbles nucleate by the clustering of He atoms and vacancies that are formed by kicking out the lattice atoms, and then grow by the migration and coalescence of helium-vacancy clusters. Small He atoms have many interstitial sites to occupy in the α-Zr HCP crystal, allowing He bubbles to grow to spherical shapes. In addition, grain boundaries induce the accumulation of primary point defects and He atoms and high irradiation temperature accelerates diffusion, and hence larger defect clusters and He bubbles form in the grain boundary regions and at higher irradiation temperature.

Insights from an OKMC simulation of dose rate effects on the irradiated microstructure of RPV model alloys

This work studies the defect features in a dilute FeMnNi alloy by an Object Kinetic Monte Carlo (OKMC) model based on the "grey-alloy" method. The dose rate effect is studied at 573 K in a wide range of dose rates from 10−8 to 10−4 displacement per atom (dpa)/s and demonstrates that the density of defect clusters rises while the average size of defect clusters decreases with increasing dose rate. However, the dose-rate effect decreases with increasing irradiation dose. The model considered two realistic mechanisms for producing <100>-type self-interstitial atom (SIA) loops and gave reasonable production ratios compared with experimental results. Our simulation shows that the proportion of <100>-type SIA loops could change obviously with the dose rate, influencing hardening prediction for various dose rates irradiation. We also investigated ways to compensate for the dose rate effect. The simulation results verified that about a 100 K temperature shift at a high dose rate of 1 × 10−4 dpa/s could produce similar irradiation microstructures to a lower dose rate of 1 × 10−7 dpa/s irradiation, including matrix defects and deduced solute migration events. The work brings new insight into the OKMC modeling and the dose rate effect of the Fe-based alloys.

Object kinetic Monte Carlo modelling of irradiation microstructures with elastic interactions

Elastic interactions between point defects and sinks, such as dislocations and cavities, affect the diffusion of point defects and are responsible for some of the features observed in microstructures under irradiation. It is therefore necessary to include elastic interactions in kinetic simulations for a quantitative prediction of material properties. In this work a method is presented to accurately and efficiently evaluate the strain field in object kinetic Monte Carlo simulations. It can handle any strain field which is biharmonic, such as the one generated by a dislocation segment or a cavity in isotropic elasticity. A speed-up of several orders of magnitude is obtained compared to the direct summation over strain sources, so that simulations over experimental time scales can be performed within reasonable computation times. The case of a thin foil containing a high density of loops under irradiation is investigated. Loop growth rates are found to depend on the loop radius, as shown experimentally, but more complex effects due to the surrounding microstructure are also highlighted.

Effects of pre-stress on the mechanical properties and microstructure of neutron-irradiated high-purity aluminum

Although structural materials in nuclear reactors are usually serviced under pre-stressed conditions, the irradiated mechanical properties and microstructures for prestressed metals have not been well understood due to the experimental challenges. In this study, the annealed high-purity aluminum was subjected to a low temperature (313 K) neutron irradiation under a stress of 1.5 MPa, which was ten percent of its unirradiated yield strength. Post-irradiation tensile tests were carried out at 300 K with a strain rate of 1.1 × 10−3 s−1. A strong radiation-induced hardening as well as a loss of uniform ductility was observed on all the specimens, and the pre-stress could defer irradiation hardening and embrittlement and decrease plastic instability stress, which was involved with a possible asymmetry distribution of voids induced by the pre-stress. Transmission electron microscopy (TEM) was conducted on the samples to quantify the void size and density as a function of pre-stress and dose. It was shown that the pre-stress could accelerate the void nucleation, increase the density but decrease the size of voids. Although previous studies have reported that the dislocation channel could be probably attributed to the micro-mechanism for irradiation embrittlement, none of dislocation channels were discovered in the deformed irradiated high-purity aluminum samples, which implied that the dislocation channel is unnecessary for irradiation embrittlement. In addition, a neutron irradiation embrittlement model for metals, which had been built on the basis of Johnson-Cook constitutive model, was found applicable to metals irradiated with pre-stress. The stimulated results also well agreed with the experimental data.

FFT-based model for irradiated aggregate microstructures in concrete

FFT-based model for irradiated aggregate microstructures in concrete

Microstructure of irradiated AGR TRISO particle buffer layers as measured by X-ray computed tomography

Shrinkage of the initially low-density buffer layer in tristructural isotropic (TRISO) coated fuel particles during irradiation is a well-known phenomenon with potential implications for fission product and actinide transport, as well as potential pyrocarbon fracture that in some cases has been observed to impact particle performance. During post-irradiation examination, the buffer layer's structure is commonly determined using 2D microscopy of a particle cross section or a series of particle cross sections at staggered depths. Although these methods provide a general idea of the irradiated buffer microstructure, they do not provide a full picture of the TRISO coating microstructure. By contrast, x-ray computed tomography (XCT) provides full 3D imaging of the TRISO particle. Particles from various compacts irradiated for the Advanced Gas Reactor Fuel Development and Qualification Program were imaged using XCT. Some of these particles were selected because of abnormal fission product inventories (e.g., low 137Cs), whereas others were randomly selected from the center of the fission product inventory distribution. Qualitative and quantitative analysis techniques were applied to the randomly selected particles as representatives of typical TRISO behavior to study the post-irradiation structure of the buffer layer. These results showed that, while separation of the buffer and inner pyrolytic carbon layers was a common behavior during the radiation-induced shrinkage of the buffer, a portion of the original buffer/inner pyrolytic carbon interface remained intact throughout irradiation in nearly all cases. These results also indicated clear trends in the degree of buffer densification with irradiation temperature and fluence.

Comparing structure-property evolution for PM-HIP and forged alloy 625 irradiated with neutrons to 1 dpa

The nuclear power industry has growing interest in qualifying powder metallurgy with hot isostatic pressing (PM-HIP) to replace traditional alloy fabrication methods for reactor structural components. But there is little known about the response of PM-HIP alloys to reactor conditions. This study directly compares the response of PM-HIP to forged Ni-base Alloy 625 under neutron irradiation doses ∼0.5–1 displacements per atom (dpa) at temperatures ranging ∼321–385 °C. Post-irradiation examination involves microstructure characterization, ASTM E8 uniaxial tensile testing, and fractography. Up through 1 dpa, PM-HIP Alloy 625 appears more resistant to irradiation-induced cavity nucleation than its forged counterpart, and consequently experiences significantly less hardening. This observed difference in performance can be explained by the higher initial dislocation density of the forged material, which represents an interstitial-biased sink that leaves a vacancy supersaturation to nucleate cavities. These findings show promise for qualification of PM-HIP Alloy 625 for nuclear applications, although higher dose studies are needed to assess the steady-state irradiated microstructure.

Complete characterization of sink-strengths for mutually 1D-mobile defect clusters: Extension to diffusion anisotropy analog cases

Simulating the long-term microstructural evolution in systems involving very fast diffusing species such as self-interstitial atom (SIA) clusters currently relies on mean-field or coarse-graining techniques. Rate-equation cluster dynamics (RECD) is one of the most popular of those when dealing with irradiated microstructure or second phase precipitation by thermal aging. Some of the most important input parameters of RECD are the absorption rates, also called cluster sink-strengths (CSS). These quantities crucially depend on the way clusters interact and diffuse and notably on the dimensionality of the involved random diffusion processes. As expected theoretically and experimentally confirmed, SIA clusters migrate in a one-dimensional fashion (possibly with random orientation changes, i.e. rotations of their Burgers vector). This complicates the calculation of the related CSS. When involving a 1D-mobile specie and an immobile reaction partner (a “1D-0” reaction) the expressions are quite well-known as well as the extension including random rotations (a “1DR-0” reaction). Expressions of CSS for absorptions between identical 1D-mobile species were proposed in the literature, but the general case of 1D-1D absorptions between different cluster classes is unknown. Here we propose a heuristic approach to such general expressions which turn out to depend on the respective capture radii of interacting clusters classes, concentrations and notably on the ratio of their respective diffusion coefficients through a power-law. In the companion paper [1], the same power-law formulation is found for 1D-3D absorptions but with different exponents, which thus appear as signatures of the dimensionality of the involved random motions. These limiting cases of CSS being established, they are finally implemented in an RECD calculation. The comparison with time-consuming kinetic Monte Carlo simulations completely validates their expression.