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The microstructural evolution of the 21Cr32Ni model alloy under ion irradiation is investigated. A set of bulk materials were irradiated at the Michigan Ion Beam Laboratory using single beam (5 MeV Fe++) to 1, 10 and 20 dpa at 440 °C and dual beam (5 MeV Fe++ plus energy degraded 1.95 MeV He++ ions) to 16.6 dpa at 446 °C. The average diameter and number density of the faulted loops and cavities formed under irradiation were characterized using Transmission Electron Microscopy (TEM). The behavior of faulted loop in the model alloy was also investigated in-situ using the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory (ANL). Results show that the average faulted loop diameter decreases, but the faulted loop number density increases with increasing dose. In-situ experiments showed that the faulted loops become unfaulted during ion irradiation by interacting with network dislocations. Although the average faulted loop diameter after 16.6 dpa dual beam irradiation at 446 °C was found to be similar to those seen in samples irradiated with single beams to 10 and 20 dpa, the faulted loop number density was significantly higher in the dual beam irradiated sample. Moreover, the dual beam irradiated model alloy exhibits a significantly higher density of smaller cavities. It is also found that the size and density of the faulted loops and voids calculated for the dual beam irradiation of 21Cr32Ni model alloy at 446 °C are in better agreement with those measured in a sample neutron irradiated at 375 °C. Further discussion is presented in this study.

This work assesses the potential of proton irradiation to simulate the neutron damage to the matrix and laves phase Zr(Fe,Cr)2 precipitates in Zircaloy-4. Isothermal proton irradiation has been performed on Zircaloy-4 samples at irradiation temperatures ranging from 250 to 350 °C. Two-step proton irradiation was also performed to enhance the amorphization of and iron loss from the laves phase Zr(Fe,Cr)2 precipitates. The irradiated microstructures, including dislocation loops and rafts near SPPs, were observed in proton irradiated Zircaloy-4, which are consistent with neutron irradiated material at a similar damage level. The amount of irradiation-induced hardening after proton irradiation was similar to post neutron irradiated data. The significant amorphization of the SPPs and concurrent Fe redistribution observed on neutron irradiated materials can be effectively emulated using a two-step proton irradiation on Zircaloy-4. Hence, the neutron irradiation effect on Zircaloy-4 can be mostly captured using the two-step proton irradiation described in this study.

Alloy 800H is a high nickel austenitic stainless steel with good high temperature mechanical properties which is considered for use in current and advanced nuclear reactor designs. The irradiation response of 800H was examined by characterizing samples that had been bulk ion irradiated at the Michigan Ion Beam Laboratory with 5 MeV Fe2+ ions to 1, 10, and 20 dpa at 440 °C. Transmission electron microscopy was used to measure the size and density of both {111} faulted dislocation loops and cavities as functions of depth from the irradiated surface. The faulted loop density increased with dose from 1 dpa up to 10 dpa where it saturated and remained approximately the same until 20 dpa. The faulted loop average diameter decreased between 1 dpa and 10 dpa and again remained approximately constant from 10 dpa to 20 dpa. Cavities were observed after irradiation doses of 10 and 20 dpa, but not after 1 dpa. The average diameter of cavities increased with dose from 10 to 20 dpa, with a corresponding small decrease in density. Cavity denuded zones were observed near the irradiated surface and near the ion implantation peak. To further understand the microstructural evolution of this alloy, FIB lift-out samples from material irradiated in bulk to 1 and 10 dpa were re-irradiated in-situ in their thin-foil geometry with 1 MeV Kr2+ ions at 440 °C at the Intermediate Voltage Electron Microscope. It was observed that the cavities formed during bulk irradiation shrank under thin-foil irradiation in-situ while dislocation loops were observed to grow and incorporate into the dislocation network. The thin-foil geometry used for in-situ irradiation is believed to cause the cavities to shrink.

The microstructural evolution of ternary Fe-21Cr-32Ni (21Cr32Ni) model alloy and of alloy 800H was investigated with a series of in-situ ion irradiation experiments performed using the Intermediate Voltage Electron Microscope (IVEM)-Tandem Facility at Argonne National Laboratory (ANL). Samples were irradiated in-situ with 1 MeV Kr++ ions in the temperature range of 50K to 713K to doses up to 2 dpa. The size distribution of defect clusters, average defect cluster diameter, and defect cluster density were measured and compared. Results showed that the evolution of defects (i.e. the average defect cluster size and number density) in 21Cr32Ni model alloy and alloy 800H with dose were similar at irradiation temperatures up to 300K where they initially increased with dose up to 0.1 dpa after which no significant changes in defect size and density were observed with further irradiation. In addition, both alloys exhibited ordered defect structures along the direction at relatively low temperatures, up to 300K, which remained stable throughout post-irradiation in-situ thermal annealing up to a temperature of 773K. During irradiation at 713K, small defect clusters were observed at low doses (

The development of TEM-visible damage in materials under irradiation at cryogenic temperatures cannot be explained using classical rate theory modeling with thermally activated reactions since at low temperatures thermal reaction rates are too low. Although point defect mobility approaches zero at low temperature, the thermal spikes induced by displacement cascades enable some atom mobility as it cools. In this work a model is developed to calculate “athermal” reaction rates from the atomic mobility within the irradiation-induced thermal spikes, including both displacement cascades and electronic stopping. The athermal reaction rates are added to a simple rate theory cluster dynamics model to allow for the simulation of microstructure evolution during irradiation at cryogenic temperatures. The rate theory model is applied to in-situ irradiation of ZrC and compares well at cryogenic temperatures. The results show that the addition of the thermal spike model makes it possible to rationalize microstructure evolution in the low temperature regime.

Irradiation grain growth of CeO2 was studied using in-situ ion irradiation with transmission electron microscopy (TEM). Thin films of ceria were produced by electron beam physical vapor deposition (EB-PVD) and then irradiated at the Intermediate Voltage Electron Microscope (IVEM) with 1 MeV Kr2+ ions at temperatures of 400°C, 600°C, and 800°C. During irradiation at the elevated temperatures, the CeO2 phase remained stable and its grains grew with irradiation. Grain growth was only weakly dependent on irradiation temperature between 400°C and 800°C. The grain growth kinetics were evaluated by a thermal spike model that was used to calculate the activation energy for grain growth.

In situ irradiations of 15Cr/15Ni Ti and 15Cr/25Ni Ti model austenitic steels were performed at the Intermediate Voltage Electron Microscope (IVEM)-Tandem user Facility (Argonne National Laboratory) at 600 °C using 1 MeV Kr++. The experiment was designed in the framework of cladding development for the GEN IV Sodium Fast Reactors (SFR). It is an extension of previous high dose irradiations on those model alloys at JANNuS-Saclay facility in France, aimed at investigating swelling mechanisms and microstructure evolution of these alloys under irradiation [1]. These studies showed a strong influence of Ni in decreasing swelling. In situ irradiations were used to continuously follow the microstructure evolution during irradiation using both diffraction contrast imaging and recording of diffraction patterns. Defect analysis, including defect size, density and nature, was performed to characterize the evolving microstructure and the swelling. Comparison of 15Cr/15Ni Ti and 15Cr/25Ni Ti irradiated microstructure has lent insight into the effect of nickel content in the development of radiation damage caused by heavy ion irradiation. The results are quantified and discussed in this paper.

Austenitic 21Cr32Ni model alloy thin foils, previously irradiated with 5 MeV Fe++ ions in bulk to create voids, were re-irradiated in-situ in the Intermediate Voltage Electron Microscope Facility (IVEM). The voids which had been formed under bulk-ion irradiation shrank and disappeared after in-situ Kr ion irradiation in the temperature range 50 K-713 K to an additional dose of 1 dpa. The voids were unaffected by eithersuccessive thermal annealing to 673 K and by prolonged exposure to the 200 keV electron beam at the irradiation temperature. The high void shrinkage rate observed did not change significantly for irradiation temperatures between 50 K and 713 K, suggesting that the void shrinkage process in thin foils during in-situ heavy-ion irradiation results from the interactions of displacement cascades with the voids. Possible void shrinkage mechanisms under thin foil irradiation are discussed in this study.

An in situ ion-irradiation study, simultaneously examined using transmission electron microscopy, was performed to investigate irradiation-induced disordering and amorphization of Al3Ti-based intermetallic compounds. Thin foil samples of two crystalline structures: D022-structured Al3Ti and L12-structured (Al,Cr)3Ti were irradiated using 1.0 MeV Kr ions at a temperature range from 40 K to 573 K to doses up to 4.06 × 1015 ions/cm2. The results showed that both the compounds underwent an order-disorder transformation under irradiation, where both Al3Ti and (Al,Cr)3Ti ordered structures were fully transformed to the disordered face-centered cubic (FCC) structure except at the highest irradiation temperature of 573 K. A slightly higher irradiation dose was required for order-disorder transformation in case of Al3Ti as compared to (Al,Cr)3Ti at a given temperature. However, their amorphization resistances were different: while the disordered FCC (Al,Cr)3Ti amorphized at the irradiation dose of 6.25 × 1014 ions/cm2 (0.92 dpa) at 40 K and 100 K, the Al3Ti compound with the same disordered FCC structure maintained crystallinity up to 4.06 × 1015 ions/cm2 (5.62 dpa) at 40 K. The critical temperature for amorphization of (Al,Cr)3Ti under Kr ion irradiation is likely between 100 K and room temperature and the critical temperature for disordering between room temperature and 573 K.

In this comparative study, in situ investigations of the microstructure evolution in a Fee9Cr ferritic emartensitic steel, NF616, and a Fee9Cre0.1C-model alloy with a similar ferriticemartensitic microstructure have been performed. NF616 and Fee9Cre0.1C-model alloy were irradiated to high doses (up to ~10 dpa) with 1 MeV Kr ions between 50 and 673 K. Defect cluster density increased with dose and saturated in both alloys. The average size of defect clusters in NF616 was constant between 50 and 573 K, on the other hand average defect size increased with dose in Fee9Cre0.1C-model alloy around ~1 dpa. At low temperatures (50e298 K), alignment of small defect clusters resulted in the formation of extensive defects in Fee9Cre0.1C-model alloy around ~2e3 dpa, while similar large defects in NF616 started to form at a high temperature of 673 K around ~5 dpa. Interaction of defect clusters with the lath boundaries were found to be much more noticeable in Fee9Cre0.1C-model alloy. Differences in the microstructural evolution of NF616 and Fee9Cre0.1C-model alloy are explained by means of the defect cluster trapping by solute atoms which depends on the solute atom concentrations in the alloys.

Zirconium carbide (ZrC) is a candidate material for use in one of the layers of TRISO coated fuel particles to be used in the Generation IV high-temperature, gas-cooled reactor, and thus it is necessary to study the effects of radiation damage on its structure. The microstructural evolution of ZrCx under irradiation was studied in situ using the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory. Samples of nominal stoichiometries ZrC0.8 and ZrC0.9 were irradiated in situ using 1 MeV Kr2+ ions at various irradiation temperatures (T = 20 K–1073 K). In situ experiments made it possible to continuously follow the evolution of the microstructure during irradiation using diffraction contrast imaging. Images and diffraction patterns were systematically recorded at selected dose points. After a threshold dose during irradiations conducted at room temperature and below, black-dot defects were observed which accumulated until saturation. Once created, the defect clusters did not move or get destroyed during irradiation so that at the final dose the low temperature microstructure consisted only of a saturation density of small defect clusters. No long-range migration of the visible defects or dynamic defect creation and elimination were observed during irradiation, but some coarsening of the microstructure with the formation of dislocation loops was observed at higher temperatures. The irradiated microstructure was found to be only weakly dependent on the stoichiometry.

Designing materials for performance in high-radiation fields can be accelerated through a carefully chosen combination of advanced multiscale modeling paired with appropriate experimental validation. The studies reported in this work, the combined efforts of six universities working together as the Consortium on Cladding and Structural Materials, use that approach to focus on improving the scientific basis for the response of ferritic–martensitic steels to irradiation. A combination of modern modeling techniques with controlled experimentation has specifically focused on improving the understanding of radiation-induced segregation, precipitate formation and growth under radiation, the stability of oxide nanoclusters, and the development of dislocation networks under radiation. Experimental studies use both model and commercial alloys, irradiated with both ion beams and neutrons. Transmission electron microscopy and atom probe are combined with both first-principles and rate theory approaches to advance the understanding of ferritic–martensitic steels.

NF616 is a nominal 9Cr ferritic–martensitic steel that is amongst the primary candidates for cladding and duct applications in the Sodium-Cooled Fast Reactor, one of the Generation IV nuclear energy systems. In this study, an in situ investigation of the microstructure evolution in NF616 under heavy ion irradiation has been conducted. NF616 was irradiated to 8.4 dpa at 50 K and to 7.6 dpa at 473 K with 1 MeV Kr ions. Nano-sized defects first appeared as white dots in dark-field TEM images and their areal density increased until saturation (∼6 dpa). Dynamic observations at 50 K and 473 K showed appearance and disappearance of TEM-visible defect clusters under irradiation that continued above saturation dose. Quantitative analysis showed no significant change in the average size (∼3–4 nm) and distribution of defect clusters with increasing dose at 50 K and 473 K. These results indicate a cascade-driven process of microstructure evolution under irradiation in these alloys that involves both the formation of TEM-visible defect clusters by various degrees of cascade overlap and cascade induced defect cluster elimination. According to this mechanism, saturation of defect cluster density is reached when the rate of defect cluster formation by overlap is equal to the rate of cluster elimination during irradiation.

In situ ion-beam irradiation was used to study irradiation induced grain growth in co-sputter-deposited Zr/xFe (0% ⩽ x ⩽ 4.5%) nanocrystalline thin films. Samples were irradiated with 500 keV Kr ions to fluences in excess of 1016 ions/cm2 (on the order of 80–100 dpa), at irradiation temperatures ranging from 20 K to 573 K. The average grain size increased monotonically with ion fluence until it reached a saturation value which depends both on temperature and on the presence of Fe. Similarly to thermal grain growth, the ion irradiation induced grain growth curves could be best fitted with curves of the type: L n - L 0 n = K Φ . Grain growth at 20 K is similar to that which occurs at 298 K. Above 298 K, the rate of grain growth increases with irradiation temperature.

Transmission electron microscopy (TEM) observations during in situ ion irradiation were used to study grain growth in free-standing Zr–Fe thin film multilayers at 25 and 300 K. Irradiations were performed with three different types of ions: 100 keV Ar, 300 keV Kr and 500 keV Xe ions to fluences of 3×10 15 ion cm −2 . Grain growth during irradiation at 20 K occurs at a similar rate to that at 300 K. At both temperatures the grain growth rate was proportional to the total number of displacements, regardless of the ion used for the irradiation. We discuss these results in terms of two previous models for grain growth under irradiation.

To further advance the mechanistic understanding of microstructural evolution in zirconium alloys for high burnup applications, it is important to obtain a quantitative measurement of the volume fractions of second-phase precipitates present in the bulk alloys as a function of the heat treatment and irradiation fluence. In this work, X-ray diffraction from a synchrotron radiation source was used to identify and follow the growth kinetics of second-phase particles in zirconium alloys. The high energy flux, energy resolution and signal-to-noise ratio of this light source allowed us to study the very small (

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...

The intermetallic compound Zr3Fe has been made amorphous by 0.9 MeV electron irradiation. By performing this irradiation in situ, it was possible to conduct a systematic study of the influence of temperature, dose rate, electron energy and specimen orientation on the amorphization process. The critical temperature and the critical dose for amorphization were determined, and shown to depend on dose rate. By varying the electron energy, we determined the displacement energies for the Zr and Fe atoms in Zr3Fe, and showed that, at low electron energy, the amorphization rate is dependent on specimen orientation. We analyze these results in terms of a model based on amorphization occurring at a damage level where the modified free energy of the irradiated crystal exceeds the free energy of the amorphous phase. This model is shown to predict the amorphization kinetics, i.e. the critical temperature and critical dose for amorphization. We also compare amorphization induced by electron and ion irradiation.

We report here a study of Kr ion implantation and resultant bubble formation in Zr and Zr alloys, including Zircaloy-2 and Zircaloy-4. Implantations into thin foils were performed in-situ in the HVEM/Tandem facility at Argonne National Laboratory at temperatures between 300 to 800°C and to doses up to 2 × 1020 ion m−2. Bulk specimens were implanted in an ion-beam chamber and then thinned for viewing by TEM. In thin foils, only small bubbles (3–10 nm) were formed at all temperatures with the exception of the Cr-rich Valloy where 13 nm bubbles were formed. Bulk samples implanted at 300°C contained a bubble morphology similar to that observed after implantation into thin foils. However at high temperatures (500–800°C) large faceted bubbles (up to 30 nm) were produced in bulk specimens. The results indicate that bubble formation and evolution below 500°C is controlled by gas concentration, while it is controlled by bubble mobility at high temperatures.

This is a review of the field of irradiation-induced amorphization of intermetallic compounds. It includes an update of recent experimental results using in-situ particle irradiation showing the effects of dose rate, temperature, crystal orientation, electron energy and the presence of stacking faults. The review describes amorphization by ion, electron and neutron irradiation in the context of a kinetic description, where the rate-limiting step is the accumulation of enough radiation damage in the lattice opposed by thermal annealing. Stability criteria, thermodynamic or otherwise, are combined with kinetics of radiation damage and annealing to provide an overall description of the amorphization process, and of the experimentally measured critical dose and critical temperature of amorphization. From the experimental observations, it is proposed that irradiation-induced amorphization in intermetallic compounds is an entropy-driven transformation, caused by the need of the material to maintain short-range order while accommodating the random ballistic motions of the atoms caused by irradiation.

The in-service degradation of reactor core materials is related to underlying changes in the irradiated microstructure. During reactor operation, structural components and cladding experience displacement of atoms by collisions with neutrons at temperatures at which the radiation-induced defects are mobile, leading to microstructure evolution under irradiation that can degrade material properties. At the doses and temperatures relevant to fast reactor operation, the microstructure evolves by dislocation loop formation and growth, microchemistry changes due to radiation-induced segregation, radiation-induced precipitation, destabilization of the existing precipitate structure, and in some cases, void formation and growth. These processes do not occur independently; rather, their evolution is highly interlinked. Radiationinduced segregation of Cr and existing chromium carbide coverage in irradiated alloy T91 track each other closely. The radiation-induced precipitation of Ni-Si precipitates and RIS of Ni and Si in alloys T91 and HCM12A are likely related. Neither the evolution of these processes nor their coupling is understood under the conditions required for materials performance in fast reactors (temperature range 300-600°C and doses beyond 200 dpa). Further, predictive modeling is not yet possible as models for microstructure evolution must be developed along with experiments to characterize these key processes and provide tools for extrapolation. To extend the range of operation of nuclear fuel cladding and structural materials in advanced nuclear energy and transmutation systems to that required for the fast reactor, the irradiation-induced evolution of the microstructure, microchemistry, and the associated mechanical properties at relevant temperatures and doses must be understood. Predictive modeling relies on an understanding of the physical processes and also on the development of microstructure and microchemical models to describe their evolution under irradiation. This project will focus on modeling microstructural and microchemical evolution of irradiated alloys by performing detailed modeling of such microstructure evolution processes coupled with well-designed in situ experiments that can provide validation and benchmarking to the computer codes. The broad scientific and technical objectives of this proposal are to evaluate the microstructure and microchemical evolution in advanced ferritic/martensitic and oxide dispersion strengthened (ODS) alloys for cladding and duct reactor materials under long-term and elevated temperature irradiation, leading to improved ability to model structural materials performance and lifetime. Specifically, we propose four research thrusts, namely Thrust 1: Identify the formation mechanism and evolution for dislocation loops with Burgers vector of a and determine whether the defect microstructure (predominately dislocation loop/dislocation density) saturates at high dose. Thrust 2: Identify whether a threshold irradiation temperature or dose exists for the nucleation of growing voids that mark the beginning of irradiation-induced swelling, and begin to probe the limits of thermal stability of the tempered Martensitic structure under irradiation. Thrust 3: Evaluate the stability of nanometer sized Y- Ti-O based oxide dispersion strengthened (ODS) particles at high fluence/temperature. Thrust 4: Evaluate the extent to which precipitates form and/or dissolve as a function of irradiation temperature and dose, and how these changes are driven by radiation induced segregation and microchemical evolutions and determined by the initial microstructure.

Zirconium carbide (ZrC) is being considered for utilization in high-temperature gas-cooled reactor fuels in deep-burn TRISO fuel. Zirconium carbide possesses a cubic B1-type crystal structure with a high melting point, exceptional hardness, and good thermal and electrical conductivities. The use of ZrC as part of the TRISO fuel requires a thorough understanding of its irradiation response. However, the radiation effects on ZrC are still poorly understood. The majority of the existing research is focused on the radiation damage phenomena at higher temperatures (>450{degree}C) where many fundamental aspects of defect production and kinetics cannot be easily distinguished. Little is known about basic defect formation, clustering, and evolution of ZrC under irradiation, although some atomistic simulation and phenomenological studies have been performed. Such detailed information is needed to construct a model describing the microstructural evolution in fast-neutron irradiated materials that will be of great technological importance for the development of ZrC-based fuel. The goal of the proposed project is to gain fundamental understanding of the radiation-induced defect formation in zirconium carbide and irradiation response by using a combination of state-of-the-art experimental methods and atomistic modeling. This project will combine (1) in situ ion irradiation at a specialized facility at a national laboratory,more » (2) controlled temperature proton irradiation on bulk samples, and (3) atomistic modeling to gain a fundamental understanding of defect formation in ZrC. The proposed project will cover the irradiation temperatures from cryogenic temperature to as high as 800{degree}C, and dose ranges from 0.1 to 100 dpa. The examination of this wide range of temperatures and doses allows us to obtain an experimental data set that can be effectively used to exercise and benchmark the computer calculations of defect properties. Combining the examination of radiation-induced microstructures mapped spatially and temporally, microstructural evolution during post-irradiation annealing, and atomistic modeling of defect formation and transport energetics will provide new, critical understanding about property changes in ZrC. The behavior of materials under irradiation is determined by the balance between damage production, defect clustering, and lattice response. In order to predict those effects at high temperatures so targeted testing can be expanded and extrapolated beyond the known database, it is necessary to determine the defect energetics and mobilities as these control damage accumulation and annealing. In particular, low-temperature irradiations are invaluable for determining the regions of defect mobility. Computer simulation techniques are particularly useful for identifying basic defect properties, especially if closely coupled with a well-constructed and complete experimental database. The close coupling of calculation and experiment in this project will provide mutual benchmarking and allow us to glean a deeper understanding of the irradiation response of ZrC, which can then be applied to the prediction of its behavior in reactor conditions.« less

Ion-beam-analysis techniques are being used to provide an understanding of the nature of collision cascades, irradiation-induced phase changes, lattice location of solute atoms and defect-solute atom interactions in various zirconium alloys. In zirconium intermetallic compounds, such as Zr3Fe, Zr2Fe, ZrFe2, and Zr3(Fex,Ni1 − x), electron and ion irradiations have been used to obtain detailed information on the crystalline-to-amorphous transformation occurring during the irradiation. Transmission-electron-microscopy (TEM) observations have provided information on the nature of the damage produced in individual cascades, the critical dose required for amorphization, and the critical temperature for amorphization. In a study on the electron-energy dependence of amorphization in Zr3Fe, Zr2Fe and ZrCr2 in situ high-voltage-electron-microscope investigations were combined with high-energy forward-elastic-recoil measurements to yield information on the threshold displacement energies for Zr and Fe or Cr in these lattices, as well as the role of secondary displacements of lattice atoms by recoil impurities (C,O) at low electron energies. In Zr implanted with 56Fe ions and subsequently bombarded with 40Ar ions at 723 K, subsequent secondary-ion-mass-spectrometry (SIMS) analyses were used to monitor the effect of irradiation on the migration of Fe in the Zr lattice. In addition, ion-channeling investigations have been used to determine the lattice sites of solute atoms in Zr as well as the details of the interaction between the solute atoms and the irradiation-produced defects.

Ion and electron irradiations were used to follow the irradiation-induced crystalline-to-amorphous transformation in Zr3Fe, ZrFe2, Zr(Cr,Fe)2 and ZrCr2, as well as in Zr(Cr,Fe)2 and Zr2(Ni,Fe) precipitates in Zircaloy-4. 40Ar and 209Bi ion irradiations of Zr3Fe were performed at 35–725 K using ions of energy 15–1500 keV. The effect of the deposited-energy density θ v in the collision cascade on the nature of the damaged regions in individual cascades was investigated. The amorphization kinetics of Zr3Fe during in situ electron irradiation were also determined. The electron fluence required for amorphization increased exponentially with temperature, and the critical temperature for amorphization was about 220 K, compared with 575–625 K for ion irradiation. The difference between the heavy ion and electron irradiation results is attributed to the fact that ion irradiation produces displacement cascades, while electron irradiation produces isolated Frenkel pairs. The dependence of the damage production of the incident electron energy was determined for Zr3Fe and the results could be analysed in terms of a composite displacement cross-section dominated at high energies by displacements of Zr and Fe atoms; by displacements of Fe atoms at intermediate energies; and by secondary displacements of lattice atoms by recoil impurities at low energies. An investigation was initiated on ZrFe2, Zr(Cr,Fe)2 and ZrCr2 to study the effect of variation of the stoichiometry and the presence of lattice defects on irradiation-induced amorphization. The irradiation-induced amorphization of the intermetallic precipitates Zr(Cr,Fe)2 and Zr2(Ni,Fe) in Zircaloy-4 was also studied during in situ bombardment by 40Ar ions of energy 350 keV. The amorphization morphology was shown to be homogeneous. These results are discussed in the context of previous experimental results of neutron and electron irradiations, and likely amorphization mechanisms are proposed.

Zirconium alloys undergo structural changes under various types of irradiation. This is particularly the case for intermetallic precipitates such as Zr2(Fe,Ni), Zr(Fe,Cr)2 in Zircaloys and Zr(Fe,V)2 in Zr-Fe-V alloy: under irradiation those phases are subject to a crystalline-to-amorphous transformation. Experimental results obtained with electron and ion irradiations are presented. With both types of irradiations, the dose to amorphization increases with the irradiation temperature and diverges when a critical temperature is reached. The relative stabilities of the four types of precipitates studied appear to be reversed for low and high irradiation temperatures.

Previous investigations using 40Ar-ion bombardments have revealed that Zr3Fe, which has an orthorhombic crystal structure, undergoes an irradiation-induced transformation from the crystalline to the amorphous state. In the present investigation, 0.9 MeV electron irradiations were performed at 28–220 K in a high-voltage electron microscope (HVEM). By measuring the onset, spread and final size of the amorphous region, factoring in the Gaussian distribution of the beam, a kinetic description of the amorphization in terms of dose, dose rate and temperature was obtained. The critical temperature for amorphization by electron irradiation was found to be ~ 220 K, compared with 570–625 K for 40Ar-ion irradiation. Also, the dose-to-amorphization increased exponentially with temperature. Results indicated that the rate of growth of the amorphous region under the electron beam decreased with increasing temperature and the dose-to-amorphization decreased with increasing dose rate. The size of the amorphous region saturated after a given dose, the final size decreasing with increasing temperature, and it is argued that this is related to the existence of a critical dose rate, which increases with temperature, and below which no amorphization occurs.

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.

Intermetallic precipitates are known to play a critical role in the oxidation process of Zircaloys. Since under irradiation they undergo structural changes, a specific study was conducted to analyze whether these transformations modify the oxidation behavior of the Zircaloy-4. Oxidation kinetics in autoclave were measured on reference, ion irradiated, and neutron irradiated materials. In the case of ion-irradiated samples, the oxidation kinetics are changed, while in the case of neutron-irradiated cladding, no significant change is observed after 60 days of oxidation. The behavior of reference and irradiated precipitates during the growth of these oxide layers was analyzed using analytical scanning transmission electron microscopy. Close to the metal-oxide interface, precipitates are incorporated unoxidized in the oxide layer. Then, when oxidized, at a few hundreds of nanometers from this interface, they undergo two major evolutions: their structure becomes either noncrystalline or occasionally amorphous and an iron redistribution and depletion is observed. In the case of precipitates previously made amorphous by irradiation, a similar behavior is observed. The role of precipitates on the oxidation of the Zircaloy-4 is discussed in terms of interaction of the precipitates with the zirconia layer (stability of the dense oxide layer) and oxidation kinetics.

The crystalline-amorphous transfornmtion of thc intermetallic precipitates Zr(Cr,Fe)z and Zr:(Ni.Fc) in Zircaloy under charged-particle and ncutron irradiation is studied. In the first section, the experimental results in the literature are reviexved for the three types of irradiation: neutron, clectron, and ion. It is found that the dose to amorphization in all three types of irradiation exhibits roughly the same exponential dependence with temperature but that there are important differences. The critical temperatures, above which amorphization is not practically attainable, are different for each type of irradiation, indicating the presence of different damage accumulation and annealing mechanisms in each case. Further exidence of this are the different amorphization morphologies observed under neutron and electron irradiation, and the shift in the relative susceptibility to amorphization between the two types of precipitate under high (neutron and ion) and low (electron and ion) temperature irradiation. In the next section, the theoretical models for amorphization are reviewed and applied to the problem in an effort to obtain a coherent picture of amorphization induced by all types of irradiation in the precipitate/zirconium system. Amorphization mechanisms are proposed for each type of irradiation, based on the experimental results. A brief conclusion indicates that different mechanisms are operative for amorphization induced by each type of irradiation and points out future areas that in our view deserve further investigation.

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 degrees C, and found to be qualitatively similar to that observed in the same alloy neutron irradiated at 420 degrees C.

We present the results of a systematic investigation of neutronirradiated and thermally annealed Fe-Cu-Ni-P model alloys using positron annihilation spectroscopy (PAS), including lifetime and Doppler broadening techniques, and Rockwell hardness. These alloys were examined in the as-fabricated state, after irradiation at 270 ~ C to 1 x 1019 n.cm "2, and to 8 x 1019 n.cm -2, and after successive postirradiation isochronal anneals at temperatures from 200 to 600 ~ C. The results can be qualitatively explained by invoking an irradiation-induced microstructure consisting of a combination of small dislocation-type defects or defect clusters (matrix damage) and dense precipitation of fine scale irradiation-induced precipitates. The matrix damage anneals between 350 ~ C and 450 ~ C. The irradiation-induced precipitates also evolve with annealing, but at higher temperatures. The combined effect of high Cu and high Ni concentrations leads to more extensive irradiation-induced precipitation than in cases where either element is missing, Whereas the effect of P is less pronounced. We analyze and compare the results with similar measurements performed on irradiated pressurevessel steels and with other positron measurements on model alloys, reported in the literature.

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.

The crystalline-amorphous transformation of the intermetallic precipitates Zr2 (Fe, Ni) and Zr(Cr, Fe)2 in Zircaloy under irradiation is studied. Experiments show that the dose-to-amorphization increases exponentially with temperature and decreases with dose rate. A model for the transformation is proposed that accounts for these observations. In this model amorphization is caused by the destabilization of the crystalline phase with respect to the amorphous phase caused by an irradiation-induced increase in its free energy. Contributions to the free energy increase due to both point defect increases and irradiation-induced-disordering are calculated and found to have approximately the same magnitude. The disordering contribution is independent of temperature and dose rate, since thermal reordering is small compared to ballistic disordering for the temperatures of interest. The temperature and dose rate dependences of the dose-to-amorphization are given by the point defect contribution. This indicates that electron-irradiation-induced amorphization is caused not only by irradiation-induced disordering but also by an increase in point defect concentration. A simplified version of the model valid at high temperature finds that the controlling parameter for amorphization is the parameter dpa.k1/2, where dpa is the dose and k the dose rate. This model is then compared with other models in the literature on the basis of amorphization kinetics and of the temperature and dose rate dependence of the dose-to-amorphization. The characteristics of the amorphous transformation under electron irradiation and neutron irradiation are discussed. It is believed that different amorphization mechanisms are operative in each case.

The overall objective of this program is to investigate the irradiation-altered phase stability of oxide precipitates in ODS steels and of model alloy solid solutions of associated systems. This information can be used to determine whether the favorable mechanical propertiies of these steels are maintained under irradiation, thus addressing one of the main materials research issues for this class of steels as identified by the GenIV working groups. The research program will also create fundamental understanding of the irradiation precipitation/dissolution problem by studying a "model" system in which the variables can be controlled and their effects understood individually.

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.

The overall aim of this project is to investigate the irradiation induced precipitation of alloying elements and dissolution of second phase particles in Zr alloys using a combination of (1) synchrotron radiation examination of bulk samples using the Advanced Photon Source (APS) at Argonne National Laboratory and (2) in-situ irradiation of model alloys using the IVEM/Tandem Facility also located at Argonne

Irradiation growth is a dimensional change at constant volume experienced by anisotropic materials such as Zr alloys when exposed to neutron irradiation. Growth rates at 350 K, although smaller than at 550 K, are significant, and models based on migration of point defects have difficulty rationalizing its characteristics. We propose an irradiation growth mechanism based on the direct athermal deposition of point defects onto extended sinks. Selfinterstitial atoms are deposited preferentially on crystallographically aligned sinks, either because of the cascade anisotropy in the Zr hcp structure, or because of their highly anisotropic diffusion through the lattice. The model is non-saturating, and the strain rate is linearly proportional to the flux as observed. Preliminary calculations show cascades are anisotropic, which can contribute to the observed growth strains.

Extended abstract of a paper presented at Microscopy and Microanalysis 2010 in Portland, Oregon, USA, August 1 – August 5, 2010.

The past ten years or so have seen the development of an exciting new class of mechanically alloyed (MA) nano-structured ferritic alloys (NFA) with outstanding mechanical properties that are mostly due to the presence of high concentrations (>10 23 m -3 ) of Ti-, Y-, and O-enriched nano-clusters (NC). Because NC may promote point defect recombination and trap transmutation-produced He in small clusters, NFA have the potential to be highly resistant to radiation damage in fission and fusion environments [1,2], and thus are being characterized following neutron and ion irradiation. Energy-filtered transmission electron microscopy (EFTEM) performed at 300 kV on a LaB6 Philips CM30 equipped with a Gatan imaging filter (GIF) has been especially beneficial for imaging NC. In particular, Fe-M jump-ratio images produced from component images recorded with 10-eV slits at energy losses of 46 and 62 eV reliably reveal NC in dark contrast. Such images are insensitive to surface oxide films or modest surface contamination and for sufficiently thin regions (

On a variety of pressure-vessel (PV) steels, we have observed changes in the average positron lifetime with increasing (near end-of-life) neutron fluences. Samples were irradiated at reactor-temperature and subjected to post-irradiation annealing, and they were examined using positron annihilation lifetime spectroscopy (PALS). The measured average positron lifetimes in high-temperature (2900 C-300° C) irradiated PV steels decrease with increasing neutron damage up to fluences of 8.5×1018 cm−2 and increase again at higher fluences. Annealing of high-fluence, 300° C irradiated ASTM A508 PV steel samples produces an initial decrease in average positron lifetimes with increasing annealing temperatures of up to 400° C, followed by an increase in average positron lifetime with higher annealing temperatures, when samples were annealed in successive 24-hour steps. A sample of weld steel, irradiated to 2.2×1019 cm−2 at 290° C, shows similar behavior in which the minimum lifetime occurs at ≈ 450° C. These trends are similar to those seen in previous studies performed on VVER and other ferritic steels.

A model is presented that estimates the effect of radiation damage on stress corrosion cracking (SCC), resulting from microstructural evolution induced by irradiation. The model is based on the Ford-Andresen film-rupture and slip-dissolution model and on the observation that the crack propagation rate increases as the material hardens. The model relates the neutron and gamma exposure to the irradiation hardening and consequent increase in the crack tip strain rate. A chemical rate theory model was employed to describe the evolution of microstructure in stainless steel leading to radiation hardening. The increase in yield stress can be correlated to an increase in the crack tip strain rate, and thus to SCC behavior. This approach was used to evaluate the effects of irradiation hardening on core shroud cracking. The results show that hardening can have a substantial effect on core shroud cracking.

A detailed study has been conducted of the effect of Kr ion irradiation on phase formation in Zr-Fe metallic multilayers, using the Intermediate Voltage Electron Microscopy (IVEM) at Argonne National Laboratory. Metallic multilayers were prepared with different overall compositions (near 50–50 and Fe-rich), and with different wavelengths (repetition thicknesses). These samples were irradiated with 300 keV Kr ions at various temperatures to investigate the final products, as well as the kinetics of phase formation. For the shorter wavelength samples, the final product was in all cases an amorphous Zr-Fe phase, in combination with Fe, while specially for the larger wavelength samples, in the Fe-rich samples the intermetallic compounds ZrFe2 and Zr3Fe were formed in addition to the amorphous phase. The dose to full reaction decreases with temperature, and with wavelength in a manner consistent with a diffusion-controlled reaction.

Vapor-deposited Zr-Fe multilayered thin films with various wavelengths and of overall composition either 50% Fe or Fe-rich up to 57 % Fe were either irradiated with 300 keV Kr ions at temperatures from 25K to 623 K to fluences up to 2 × 1016 cm−2, or simply annealed at 773K in-situ in the Intermediate Voltage Electron Microscope at Argonne National Laboratory. Under irradiation, the final reaction product is the amorphous phase in all cases studied, but the dose to amorphization depends on the temperature and on the wavelength. In the purely thermal case (annealing at 773 K), the 50–50 composition produces the amorphous phase but for the Fe-rich multilayers the reaction products depend on the multilayer wavelength. For small wavelength, the amorphous phase is still formed, but at large wavelength the Zr-Fe crystalline intermetallic compounds appear. These results are discussed in terms of existing models of irradiation kinetics and phase selection during solid state reaction.

The Zr-based intermetallic compounds ZrCr2, ZrFe2 and Zr3(Fei_x,Nix) were irradiated with high energy electrons at the HVEM/Tandem facility at Argonne National Laboratory to study their amorphization behavior. The results show that although ZrCr2 and ZrFe2 have the same Laves phase C15 fee crystal structure, their critical temperatures for amorphization under electron irradiation were 180 K and 80 K, showing that the substitution of Cr for Fe in the sublattice had a marked effect on the annealing characteristics of the material. The low temperature dose to amorphization was higher in ZrFe2 than in ZrCr2 by a factor of two. The presence of a high density of stacking faults had a strong effect on amorphization in both compounds causing the critical temperature to be increased by 10–15 K. By contrast, the addition of Ni to Zr3(Fei_x,Nix) had no effect on amorphization behavior for x=0. 1 and 0. 5. These results are discussed in terms of current models of amorphization based on defect accumulation and the attainment of a critical damage level, such as given by the Lindemann criterion.

We report here the results of a study conducted to examine the effect of Kr ion irradiation on bubble formation in Zr alloys. We used the HVEM/Tandem facility at Argonne National Laboratory to irradiate several Zr alloys, including Zircaloy-2 and Zircaloy-4, at temperatures from 300 to 800 C and to doses up to 2 × 1016ion.cm−2. Both in-situ irradiation of thin foils as well as irradiation of bulk samples with an ion implanter were used in this study. For the thin foil irradiations, a distribution of small bubbles in the range of 30-100 Å was found, at all temperatures with the exception of the Cr-rich Valloy where bubbles of 130 Å were found. The irradiation of bulk samples at high temperature (700–800 C) produced large faceted bubbles (up to 300 Å) after irradiation to 2 × 1016ion.cm−2. The results are examined in the context of existing models for bubble formation and growth in other metals.

The binary and ternary intermetallic compounds Zr3Fe, Zr2 Fe, (Zr0.5,Nb0.5)3Fe, Zr3(Fe0.9,Ni0.1) and Zr3(Fe0.5,Ni0.5) were subjected to 900 keV electron irradiation until amorphous to study the change in the dose-to-amorphization with temperature. The critical temperatures were observed to vary with dose rate, and with the type of compound. Hexagonal (Zr0.5,Nb0.5)3Fe had an appreciably lower critical temperature and higher dose to amorphization at low temperature than orthorombic Zr3Fe, whereas other orthorombic Zr3(Fex,NiI-x) compounds were essentially identical in behavior to Zr3Fe. The electron energy dependence of the dose-to-amorphization was studied in Zr3Fe between 250 and 900 keV. The analysis of the results gives displacement energies of EZrd = 26 eV, EFed = 18 eV in the Zr3Fe compound.