Download or read book Irradiation Induced Growth and Microstructure Evolution of Zr 1 2Sn 1Nb 0 4Fe Under Neutron Irradiation to High Doses written by GP. Kobylyansky and published by . This book was released on 1999 with total page 17 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zirconium alloy components subjected to long-term operation and high doses in thermal reactors need to be highly irradiation resistant to provide integrity of components, primarily, their geometrical sizes. Transmission and scanning electron microscopy, energy dispersive X-ray microanalysis used to investigate thin foils and extraction replicas of irradiated zirconium, Zr-1Nb (E110) and Zr-1.2Sn-1Nb-0.4Fe (E635) alloys allowed us to analyze the evolution of their microstructure under neutron irradiation. The experimental irradiations that were conducted at 350°C to 1027 n/m2 (E >= 0.1 MeV) show that the most irradiation resistant alloy proved to be a multicomponent E635 alloy. It is not essentially subject to growth. Dislocation structure and phase composition were studied as interrelated to different stages of irradiation induced growth. The accelerated growth correlates with a high density of basal -- plane c-component dislocations.

Download or read book Irradiation Creep and Growth Behavior and Microstructural Evolution of Advanced Zr Base Alloys written by A. Soniak and published by . This book was released on 2000 with total page 23 pages. Available in PDF, EPUB and Kindle. Book excerpt: This paper deals with the irradiation-induced changes in the microstructure of FRAMATOME advanced Zr-base alloys and the correlations with their irradiation creep and growth behavior. The first part is dedicated to experimental irradiations performed at 280 and 350°C in a CEA metallurgical test reactor (Siloé, 1 to 1018 n/m2 s, E > 1 MeV) on stress-relieved (SRA) and recrystallized (RXA) low-tin Zircaloy-4 and two advanced RXA materials (Alloy 4 (M4): Zr-SnFeV, and Alloy 5 (M5): Zr-NbO), which are proposed for fuel rod cladding applications in PWRs. The irradiation creep results confirm the improved behavior of RXA Zy-4, M4, and M5 in comparison to that of SRA Zy-4. Similar trends are observed for irradiation growth, the SRA Zy-4 exhibiting a quasi-linear behavior with increasing fluence while RXA alloys undergo an early saturation phenomenon. Among RXA materials, M5 has the higher irradiation growth resistance. These creep and growth results at moderate neutron fluences (

Download or read book Microstructure Evolution in Zr Alloys During Irradiation written by M. Griffiths and published by . This book was released on 2008 with total page 8 pages. Available in PDF, EPUB and Kindle. Book excerpt: The performance of zirconium alloys in BWR, PWR, and PHWR nuclear reactors is dependent on the microstructure. Accordingly, the characterization of the microstructure is an integral part of any study conducted to develop models for in-reactor performance. Although the as-fabricated microstructure (texture, grain size, dislocation density, and phase or precipitate distribution) determines the basic physical properties of a given component, there are changes that occur during irradiation that can have a significant effect on these properties. Microstructures that illustrate specific features of the radiation damage that forms in Zr alloys will be illustrated and discussed in terms of the dose, dose rate, and impurity factors that are applicable.

Download or read book Irradiation Induced Growth and Microstructure of Recrystallized Cold Worked and Quenched Zircaloy 2 NSF and E635 Alloys written by D. W. White and published by . This book was released on 2008 with total page 19 pages. Available in PDF, EPUB and Kindle. Book excerpt: This paper is devoted to the study of the effect of the texture, phase composition, and microstructure on the irradiation-induced growth strain (GS) of zirconium-based alloys. GS measurements and TEM microstructural examinations were performed on Zry-2, NSF, and E635 samples in the recrystallized, beta quenched and cold-worked (CW) conditions. The samples were irradiated in the BOR-60 reactor in the temperature range of 315-325°C up to a neutron fluence level of 1.1 x 1026 n/m2 (E>1MeV), i.e., up to a damage dose of 23 dpa. Growth strains of NSF and E635 alloys in all states and in the longitudinal and transverse directions are lower as compared to those of Zry-2, and do not exceed 0.2 % even at the maximum fluence level. As for recrystallized Zry-2, the GS kinetics are characterized by the appearance of the accelerated growth stage. A combination of a certain amount of Nb, Fe, and Sn in the matrix content plays a key role in GS kinetics. The higher the degree of CW, the higher the irradiation growth but its rate of increase with increasing fluence is different for alloys of different compositions. The maximum GS, reaching 0.72 %, is observed in the 20 % CW Zry-2 samples. Texture, along with the alloy composition, is one of the main GS-determining factors. Irradiation growth of the transversal samples is lower as compared to the longitudinal ones because of texture. As for quenched alloys, the texture is practically isotropic and GS values are low, independent of the alloy composition. In CW materials, the density of ‹c›- dislocations greatly affects the irradiation growth strain. Particles of Zr(Fe,Cr)2 and Zr2(Fe,Ni) phases in Zry-2 as well as Zr(Nb,Fe)2 in NSF and E635 are depleted in iron under irradiation. The Fe goes into the matrix and modifies its properties. The HCP lattice structure in the Laves phases in NSF and E635 changes into BCC (?-Nb-type). FCC (Zr,Nb)2Fe precipitates preserve on the whole their composition and structure; no amorphization of the Nb-containing precipitates is observed. The Zr2(Fe,Ni) precipitates with a BCT lattice remain crystalline, and HCP Zr(Cr,Fe)2 precipitates undergo amorphization. The average particle size in the irradiated alloys is larger and the concentration is a little lower as compared to the unirradiated ones. Irradiation-induced fine dispersed precipitates about 3 nm in size, probably enriched in niobium, appear in NSF and E635. The observed changes of microhardness are discussed from the viewpoint of generation of radiation defects (clusters, dislocation loops), evolution of the initial dislocation structure, and matrix composition (enrichment in Fe, Cr, and, probably, Nb).

Download or read book Microstructure Evolution of Zirconium Carbide Irradiated by Ions written by Christopher Ulmer and published by . This book was released on 2014 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: ZrC is a candidate material for use in Generation IV high-temperature, gas-cooled reactor TRISO coated fuel particles, so it is important to understand its behavior under irradiation. The microstructural evolution of ZrC$_x$ under irradiation was studied in situ using the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory. Experiments were performed in which the sample stoichiometry and irradiation temperature were systematically varied. In situ experiments made it possible to continuously follow the microstructure during irradiation using diffraction contrast imaging. Images and diffraction patterns were methodically recorded at chosen dose points. Experiments centered on the irradiation of ZrC$_{0.8}$ and ZrC$_{0.9}$ with 1 MeV Kr ions at temperatures ranging from 20 - 1073 K up to 10 dpa.Initial damage developed as 2 - 4 nm diameter black-dot defects after a threshold dose of approximately 0.1 - 0.5 dpa. As the irradiation temperature increased, the threshold dose for visible defect formation decreased. The density and size of defects increased with additional dose and the density of defects ranged on the order of $10^{22}$ - $10^{23}$ m$^{-3}$ for all experiments. The defect diameter also increased with irradiation temperature, with average defect diameters at 3 dpa ranging from approximately 4 nm at 673 K to 8 nm at 1073 K. No long-range migration of the visible defects or dynamic defect creation and elimination were observed during irradiation, but agglomeration of small defects into loops occurred at 1073 K and resulted in an overall coarsening of the microstructure. The irradiated microstructure was found to not be strongly dependent on the stoichiometry as results for the two stoichiometries studied were nearly identical. No irradiation induced amorphization was observed, even after 5 dpa at 20 K and 10 dpa at 50 K. At the higher temperature (873 K and above), the irradiated microstructure varied with sample thickness and showed a defect-denuded zone in the thin area near the edge.A one-dimensional cluster dynamics rate theory model that only considered the creation and mobility of point defects and their agglomeration into defect clusters was solved and compared with the experimental results. General trends from the simulation results matched the experimental observations: a threshold dose was predicted by the calculation, loop diameter was predicted to increases with dose and temperature, and loop density increased with dose and decreased with temperature, as observed. The spatial distribution showed lower loop size and density near the surface. Additional work is needed to match the experimental results quantitatively for both loop size and density, and the results were found to be sensitive to the chosen temperature.

Download or read book Influence of Zirconium Alloy Chemical Composition on Microstructure Formation and Irradiation Induced Growth written by AV. Tselischev and published by . This book was released on 2002 with total page 22 pages. Available in PDF, EPUB and Kindle. Book excerpt: The studies of the dislocation structure, phase, and microchemical compositions of alloy Zr-1Nb-1.2Sn-0.35Fe (E635) and its modifications containing Fe from 0.15 to 0.65% were carried out before and after research reactor irradiation at ~350°C to maximal fluence of ~1027 m-2 (E > 0.1 MeV) and at ~60°C. The size and concentration of the a-type loops depend on the alloy composition and fluence and saturate even at low doses (1 dpa). The evolution of the c-component dislocation structure in recrystallized alloys of E365 type is determined by the chemical and phase compositions of alloys specifically, by the Fe/Nb ratio and the threshold dose, and is consistent with the irradiation growth strain acceleration. In E635 alloy containing 0.15%Fe the accelerated growth is observed after the dose of 15 dpa and is attended with the evolution of the c dislocation structure which is similar to Zr-1Nb (E110) alloy behavior. The irradiation induced growth of E635 type alloy containing 0.65% Fe is similar to that of E635 having the normal composition; no

Download or read book Influence of Structure Phase State of Nb Containing Zr Alloys on Irradiation Induced Growth written by VN. Shishov and published by . This book was released on 2005 with total page 20 pages. Available in PDF, EPUB and Kindle. Book excerpt: On account of the search for the optimal composition and structure-phase state of Zr alloys much attention is paid to upgrade the E110 (Zr-1 %Nb) and E635 (Zr-1 %Nb-0.35 %Fe-1.2 %Sn) alloys that have proved well in terms of irradiation-induced creep and growth, high strength characteristics, and corrosion. The difference between the alloy properties is determined by their states related to their compositions. The structure-phase state of the Zr-Nb and Zr-Nb-Fe-Sn systems has been studied after heat treatment in the ?-- and ? + ?- regions and its influence on the irradiation-induced growth (IIG) during BOR-60 irradiation at T =315-350 %C was investigated. A substantial difference has been shown in the deformation effected by IIG of those alloys; it is less for Zr-Nb-Fe-Sn alloys in dissimilar structure-phase states. The incubation period of the accelerated growth stage is determined by the ?-matrix composition, the phase state and the initial dislocation structure. Neutron irradiation leads to a redistribution of alloying elements between the matrix and the precipitates, and to changes in the ?-solid solution composition. These changes affect accumulation and mobility of irradiation defects, anisotropy and formation of vacancy c-component dislocation loops. The appearance of c-loops usually correlates with an axial direction acceleration of the IIG of tubes conforming to their texture. The basic regularities of the phase transformation have been established: a) ?-Nb precipitates in Zr-Nb alloys are altered in composition to reduce the Nb content from 85-90 % to ~ 50 %, fine precipitates likely enriched in Nb are formed; b) ?-Zr precipitates are subject to irradiation-stimulated decomposition; c) Laves phase precipitates change composition (the content of Fe decreases) and crystal structure, HCP to BCC (?-Nb); d) (Zr,Nb)2Fe precipitates having the FCC lattice retain their composition and crystal structure; e) no amorphization of any secondary phase precipitates is observable under the given conditions of irradiation (T = 315-350 °C). Based on the dpa, the results were compared pertaining to Zr-alloy IIG deformation vs. fluence in various reactors at different energies of fast neutrons. The presented graphs enable comparison between the results of numerous experiments and enable predictions of Zr-material behavior in long-term operation and at high burn-up in commercial reactors.

Download or read book Microstructural Development in Neutron Irradiated Zircaloy 4 written by WJS Yang and published by . This book was released on 1990 with total page 15 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zircaloy-4, a zirconium base alloy used extensively as cladding and core structural material in water cooled nuclear reactors, was examined by transmission electron microscopy after neutron irradiation and postirradiation annealing. Phase instabilities found during irradiation include the amorphous transformation and the dissolution of intermetallic precipitate Zr(Fe,Cr)2 in the ?-recrystallized matrix and the dissolution of the metastable precipitate Zr4(Fe,Cr) in the ?-quenched matrix. The alloy is driven toward a single phase solid solution during the irradiation. The presence of fast diffusion iron species in the matrix due to the precipitate dissolution may have caused the irradiation growth breakaway phenomenon. The microstructural evolution during irradiation consists of ̄c dislocation development and grain boundary migration. The presence of ̄c dislocations indicates permanent strain in the matrix. The postirradiation annealing at 833 K does not anneal out the ̄c dislocations. The ̄c dislocation is postulated to have developed due to the intergranular constraints under the continuous growth in the breakaway region.

Download or read book Theoretical Investigation of Microstructure Evolution and Deformation of Zirconium Under Neutron Irradiation written by and published by . This book was released on 2015 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: We studied the radiation growth of zirconium using a reaction-diffusion model which takes into account intra-cascade clustering of self-interstitial atoms and one-dimensional diffusion of interstitial clusters. The observed dose dependence of strain rates is accounted for by accumulation of sessile dislocation loops during irradiation. Moreover, the computational model developed and fitted to available experimental data is applied to study deformation of Zr single crystals under irradiation up to hundred dpa. Finally, the effect of cold work and the reasons for negative prismatic strains and co-existence of vacancy and interstitial loops are elucidated.

Download or read book Peculiarities of Structural and Behavioral Changes of Some Zirconium Alloys at Damage Doses of Up to 50 Dpa written by VN. Shishov and published by . This book was released on 2004 with total page 14 pages. Available in PDF, EPUB and Kindle. Book excerpt: The irradiation-induced damage of zirconium alloys subjected to neutron irradiation up to dose levels of ~50 dpa was investigated. Specimens of unalloyed zirconium, Zr-1%Nb, Zr-2.5%Nb and Zr-1%Nb-1.3%Sn-0.4%Fe were irradiated in the BOR-60 reactor over the temperature range 320-420°C. The dose dependence of the irradiation growth strain increased sharply in zirconium and Zr-Nb irradiated at ~350°C at doses above ~10 dpa. In the case of Zr-1%Nb-1.3%Sn-0.4%Fe, it increased at doses of ~37 dpa. Upon increasing the irradiation temperature to 420°C, a sharp accelerated irradiation growth of the Zr-1%Nb alloy began shifting up to about 30 dpa. For the Zr- 1%Nb-1.3%Sn-0.4%Fe, no change of the irradiation growth rate was observed up to a dose of 55 dpa. The onset of increased irradiation growth in alloys correlates with the occurrence of c-component dislocation loops which coincides with a loss of coherence of finely-dispersed precipitates. Post-irradiation annealing experiments demonstrated that a delay in loop formation leads to displacement of the "break-away" beginning in the dose dependence of the irradiation growth in the direction of high doses. The a+c-type dislocation loops were also formed in Zr-1%Nb alloy at high doses, but their influence on the change of macroscopic properties was not observed.

Download or read book Irradiation Induced Redstribution of Alloying Elements in Zr Nb Alloys and Its Effect on Corrosion Kinetics written by Zefeng Yu and published by . This book was released on 2020 with total page 276 pages. Available in PDF, EPUB and Kindle. Book excerpt: Zirconium-based alloys have been used in nuclear fission reactors, because of their low thermal neutron cross-section, good mechanical strength, and adequate corrosion resistance. In pressurized water reactors, one of the major reasons that Zr-Nb alloys have widely replaced Zircaloys-4 is the absence of the accelerated oxide growth at high burnup. Although such great advantages have led to the development of many advanced commercial Zr-Nb alloys, the reasons behind the enhanced in-reactor corrosion resistance are still unclear. The distribution and concentration of alloying elements in the substrate have been suspected bo play a major role in controlling in-reactor corrosion kinetics. The major hypothesis being tested in this thesis study is that the enhanced corrosion resistance of irradiated Zr-Nb alloys is a result of irradiation-induced reduction of Nb concentration in solid solution, due to the nucleation and growth of Nb-rich irradiation-induced platelets (IIPs)/nanoclusters. To validate our hypothesis, a systematic study has been performed to understand the irradiation induced microstructure and microchemistry evolution and the subsequent effect on the corrosion kinetics of Zr-xNb (x=0.2, 0,4, 0,5, 1.0) model alloys. The microstructure and microchemistry of as-received materials were characterized under STEM/EDS/APT. Then, 2 MeV proton irradiation has been performed on these model alloys at 350 °C up to 1 dpa. (S)TEM/EDS has been used to study the size and density evolution of the native precipitate and IIPs as a function of irradiation dose. The major use of APT is to quantify the Nb concentration in the solid solution as a function of irradiation dose in order to support our hypothesis. The IIPs crystal structure and growth mechanism have been particularly inverstigated using HRSTEM and 4D-STEM. After the characterization, the irradiated materials were corroded in autoclave to study if the proton irradiation leads to subsequent lower corrosion rate. Lastly, the same characterization techniques and methods have been used to study neutron irradiated commercial alloys, M5®, ZIRLO® and X2®, in an effort to compare the results with proton irradiation. The possible IIPs nucleation and growth mechanism and the effects of irradiation-induced Nb redistribution on the corrosion kinetics are the major focuses of the discussion section.

Download or read book Evolution of Dislocation and Precipitate Structure in Zr Alloys Under Long Term Irradiation written by LP. Sinelnikov and published by . This book was released on 2000 with total page 17 pages. Available in PDF, EPUB and Kindle. Book excerpt: Tubes from zirconium-base alloys are used widely in the pressure tube reactor core. The lifetime of the zirconium component in the reactor core will be determined by structure changes and alloy properties under long-term neutron irradiation. The studies were carried out using Zr-1Sn-1Nb-0.4Fe (E635) and Zr-2.5Nb (E125) alloy samples cut out of a pressure tube (PT) in the initial condition and after 7 and 15.5 years operated (42 000 and 95 000 effective hours) under irradiation to the neutron fluxes of 3 x 1017 and 2 x 1017 n/m2 s (E > 1 MeV) at 304°C in RBMK-1000 and 314°C in RBMK-1500, respectively. The E125 alloy PTs were in two conditions, as cold worked and annealed (A) and as thermomechanically treated (TMT-1) (B). The E635 alloy PTs were cold worked and annealed (A) (Tablel). The examinations were implemented using analytical transmission electron microscopy (TEM), energy dispersive X-ray (EDX), and X-ray diffraction (XRD) analyses. New data showing the microstructure changes are presented. Both the alloys have a partially recrystallized grain structure with a high density of intragranular dislocations in the initial state. The main part of dislocations belong to ?a? type. Density of secondary phase precipitates is high. They are ?-Nb (bec) in Zr-2.5Nb. In Zr-1.3Sn-1Nb-0.4Fe, precipitates consist of Zr, Nb, and Fe, and the constituent ratio is close to 1:1:1 Zr(Nb,Fe)2 (hcp). Linear dislocations (Type a) are annealed under irradiation, while the density of ?c?-component dislocations is not practically changed. Grain structure of the Zr-2.5Nb alloy is retained, and it is practically completely recrystallized in Zr-1.3Sn-1Nb-0.4Fe. The phase structure and microchemical composition are modified by irradiation. Nb concentration changes in ?-Nb are observed in Zr-2.5Nb. A substantial decrease of Fe concentration and irradiation defect accumulation are observed in the intermetallic precipitates Zr(Nb,Fe)2 in the E635 alloy. This leads to crystal lattice disordering and new precipitates Nb-enriched are formed. Dislocation loops are formed under irradiation. Loop dimensions vary widely in Zr-2.5Nb. They show a tendency to ordering under high-fluence irradiation. Uniform structure of loops with a high tendency to ordering is formed in the alloy Zr-l.3Sn-lNb-0.4Fe; 70% of them are interstitial loops of the ?a? type. Irradiation-induced Fe depletion of intermetallic particles and a Fe content increase in saturated ?-Zr matrix may be a cause of the microstructure and performance changes in E635 alloy pressure tubes. The correlation between irradiation-induced dislocation structure and hardening of the E125 alloy is discussed.

Download or read book THEORETICAL INVESTIGATION OF MICROSTRUCTURE EVOLUTION AND DEFORMATION OF ZIRCONIUM UNDER CASCADE DAMAGE CONDITIONS written by and published by . This book was released on 2012 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This work is based on our reaction-diffusion model of radiation growth of Zr-based materials proposed recently in [1]. In [1], the equations for the strain rates in unloaded pure crystal under cascade damage conditions of, e.g., neutron or heavy-ion irradiation were derived as functions of dislocation densities, which include contributions from dislocation loops, and spatial distribution of their Burgers vectors. The model takes into account the intra-cascade clustering of self-interstitial atoms and their one-dimensional diffusion; explains the growth stages, including the break-away growth of pre-annealed samples; and accounts for some striking observations, such as of negative strain in prismatic direction, and co-existence of vacancy- and interstitial-type prismatic loops. In this report, the change of dislocation densities due to accumulation of sessile dislocation loops is taken into account explicitly to investigate the dose dependence of radiation growth. The dose dependence of climb rates of dislocations is calculated, which is important for the climb-induced glide model of radiation creep. The results of fitting the model to available experimental data and some numerical calculations of the strain behavior of Zr for different initial dislocation structures are presented and discussed. The computer code RIMD-ZR. V1 (Radiation Induced Microstructure and Deformation of Zr) developed is described and attached to this report.

Download or read book Irradiation Growth of Zirconium Alloys written by JY. Ren and published by . This book was released on 1994 with total page 8 pages. Available in PDF, EPUB and Kindle. Book excerpt: Experimental investigation of irradiation growth on annealed Zircaloy-4 and 20% to 50% cold-worked Zr-2.5wt%Nb specimens with stress relief has been carried out. The specimens are irradiated in a heavy water reactor at 610 K to 4.2 x 1024 n/m2 (E > 1.0 MeV). The growth strains increase linearly with fluence. The saturation of growth is not observed for all specimens. The difference of growth behavior between two kinds of Zircaloy-4 tube may be associated with the content of minor alloying elements and impurities that influence the microstructure evolution under irradiation.

Download or read book Evolution of Microstructure in Zirconium Alloys During Irradiation written by M. Griffiths and published by . This book was released on 1996 with total page 23 pages. Available in PDF, EPUB and Kindle. Book excerpt: X-ray diffraction (XRD) and transmission electron microscopy (TEM) have been used to characterize microstructural and microchemical changes produced by neutron irradiation in zirconium and zirconium alloys. Zircaloy-2, Zircaloy-4, and Zr-2.5Nb alloys with differing metallurgical states have been analyzed after irradiation for neutron fluences up to 25 x 1025 n.m-2 (E > 1 MeV) for a range of temperatures between 330 and 580 K.

Download or read book Evolution of Microstructure in Zirconium Alloy Core Components of Nuclear Reactors During Service written by PCK Chow and published by . This book was released on 1994 with total page 34 pages. Available in PDF, EPUB and Kindle. Book excerpt: X-ray diffraction (XRD) and analytical electron microscopy (AEM) have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components.

Download or read book Influence of Neutron Irradiation on Dislocation Structure and Phase Composition of Zr Base Alloys written by VN. Shishov and published by . This book was released on 1996 with total page 20 pages. Available in PDF, EPUB and Kindle. Book excerpt: Studied were evolution of dislocation structure, phase, and element composition of binary alloys Zr-1Nb and Zr-2.5Nb and multicomponent alloys Zr-1Nb-1.2Sn-0.4Fe and Zr-1.2Sn-0.4Fe under neutron irradiation. The investigations were carried out using cladding and pressure tubes before and after irradiation to a fluence of ~1026 n/m2 (E >= 0.1 MeV) in experimental and commercial reactors at 300 to 350°C using TEM, EDX, and XRD. In most cases, irradiation-induced defects are in the form of dislocation loops with Burgers vector 1/3 ?1120?. The density of dislocations with a ?c? component is less than 2 x 1014 m-2. A higher fluence or the presence of strain results in the ordering of the dislocation structure of ?c? component and ?a?-type dislocation loops. Before irradiation, the multicomponent alloys contain fine precipitates of Zr-Nb-Fe composition, and the matrix is depleted in Fe. Under irradiation, recrystallization proceeds intensively (as distinct from Zr-Nb alloys), changes take place in size, distribution, and composition of precipitates (with a relative decrease of Fe content compared to Nb), and the Fecontent of ?-Zr matrix is increased. None of the materials studied showed any significant evidence of secondary phase particle amorphization. The density of dislocations with ?a? and ?c? components and irradiation-induced defects, their mean size, the extent of ordering, and the planes of their occurrence were determined. A comparison was made between irradiation-induced evolutions of microstructures of the different alloys.