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Nanostructural evolution of Cr (Cr-rich) precipitates in a Cu-0.78%Cr-0.13%Zr alloy has been studied after aging and overaging (reaging) by laser assisted local electrode 3 dimensional atom probe (Laser-LEAP). This material is a candidate for the first wall and divertor components of future fusion reactors. After prime aging at 460 C, Cr precipitates enriched with Zr were observed. Further reaging at 600 C caused the precipitates to grow to almost spherical Cr precipitates with 5 nm (1 h) in diameter and plate-like ones with 20 nm (4 h), respectively. Zr and impurities of Si and Fe were concentrated around the Cr precipitates, resulting in an almost pure Cr cores with interface regions enriched with Zr and the impurities. Probably, the strain induced by the incoherency of BCC Cr with matrix FCC Cu is relaxed by the formation of the enriched regions. [doi:10.2320/matertrans.MBW200736]

We review the current status of the European fusion materials modelling programme. We describe recent findings and outline potential areas for future development. Large-scale density functional theory (DFT) calculations reveal the structure of the point defects in α-Fe, and highlight the crucial part played by magnetism. The calculations give accurate migration energies of point defects and the strength of their interaction with He atoms. Kinetic models based on DFT results reproduce the stages of radiation damage recovery in iron, and stages of He-desorption from pre-implanted iron. Experiments aimed at validating the models will be carried out in the future using a multi-beam ion irradiation facility chosen for its versatility and rapid feedback.

At a low dose level of 0.001 dpa the Ti6Al4V (α + β) alloy showed softening and at higher doses an increase in hardening and lack of work hardening after irradiation at 60 °C. The (α + β) alloy seemed to suffer from plastic instability when irradiated to a dose level of 0.3 dpa at 60 °C. At elevated temperatures a substantial amount of hardening was observed in the (α + β) alloy when tested in the irradiated condition. Earlier studies have shown that the fracture toughness behaviour of the irradiated Ti5Al2.5Sn (α) and (α + β) alloys were quite similar at ambient temperatures. At elevated temperatures, fracture toughness of the irradiated (α + β) alloy decreased more than that of the (α) alloy when compared to the results obtained in the unirradiated condition. The large irradiation hardening and loss of fracture toughness in the (α + β) alloy appears to be related to radiation dose level, temperature and radiation-induced precipitation in the (α + β) alloy.

The fatigue and creep–fatigue response of copper alloys is of interest due to the cyclic thermal–mechanical loading processes a fusion first wall will experience during operation. Creep–fatigue experiments were performed on a CuCrZr alloy with an overaged heat treatment at room temperature to determine the effects on fatigue life of a 10 s hold period applied at the maximum tension and compression points in the fatigue loading cycle. The hold period produced a reduction in the number of cycles to failure. This reduction was largest at the lowest strain amplitudes and the longest fatigue lives, the region of most interest for component design. Stress relaxation was observed during the hold periods even at room temperature where thermally-activated creep processes are not expected. The large reduction in fatigue life is apparently due to a change in the crack initiation mode from transgranular with no hold period to intergranular with a hold period.

The precipitate microstructure of prime aged CuCrZr was coarsened by overaging to see if the larger precipitates could prevent the initiation of plastic flow localization in irradiated CuCrZr. A number of tensile and fracture toughness specimens of prime aged CuCrZr alloy were given overaging treatments at 873 K for 1, 2 and 4 h. A subset of these specimens were irradiated with fission neutrons in the BR-2 reactor at Mol (Belgium) at 333 K and 573 K to a dose level of ∼0.3 dpa. Both unirradiated and irradiated tensile specimens were tested at 295 K, 333 K and 573 K. Fracture toughness tests were carried out at 293, 333 and 573 K. Transmission electron microscopy was used to investigate the effects of overaging, subsequent irradiation and the effect of deformation. The results indicate that the overaging treatment of 873 K for 1 h produced a precipitate microstructure that improved the plastic instability, overall ductility and fracture toughness.

The reaction kinetics of the various species of mobile defects in irradiated materials are crucially dependent on the dimensionality of their migration. Sink strengths for one-dimensionally (1D) gliding interstitial loops undergoing occasional direction changes have been described analytically and confirmed by kinetic Monte Carlo (KMC) simulations. Here we report on KMC simulations investigating a different transition from 1D to 3D diffusion of 1D gliding loops for which their 1D migration is interrupted by occasional 2D migration due to conservative climb by dislocation core diffusion within a plane transverse to their 1D glide direction. Their transition from 1D to 3D kinetics is significantly different from that due to direction changes. The KMC results are compared to an analytical description of this diffusion mode in the form of a master curve relating the 1D normalized sink strength to the frequency of disturbance of 1D migration.

Worldwide programs aimed at developing materials for future fusion plants are presently concentrating on reduced activation ferritic/martensitic (RAFM) steels. In Europe, a Fusion Long Term Programme is being carried out under the coordination of European Fusion Development Agreement (EFDA) Close Support Unit (Garching, Germany); in the framework of this programme, extensive research is currently in progress for the development and full qualification of the European reference 9Cr RAFM steel, designated EUROFER. This paper focuses on the work being performed within the EFDA Technology Workprogram TTMS-001 task "Irradiation Perfonnance of EUROFER", aimed at defining the post-irradiation mechanical and microstructural characteristics of the material, through irradiations, post-irradiation examinations (PlEs) and theoretical studies. An overview of the progress made will be given, as well as an anticipation of future research activities. (c) 2005 Elsevier B.V. All rights reserved.

The formation of ‘cleared’ channels in neutron irradiated metals and alloys have been frequently reported for more than 40 years. So far, however, no unambiguous and conclusive evidence showing as to how and where these channels are initiated has emerged. In the following we present experimental results illustrating initiation and propagation of channels during post-irradiation deformation of neutron irradiated copper and a copper alloy. The observations strongly suggest that the channels are initiated at boundaries, large inclusions and even at previously formed cleared channels. Some of the channels have been observed to penetrate through both the twin boundaries and grain boundaries. It is argued that the high stress level reached during post-irradiation tensile tests activate dislocation sources at the sites of stress concentrations at boundaries and interfaces. The propagation of these newly generated dislocations in the matrix causes the formation of cleared channels. Implications of these results are discussed with specific reference to the origin and consequences of plastic flow localization.

Clusters of self-interstitial atoms (SIAs) in the form of parallel crowdions are created directly in high-energy displacement cascades produced in metals by neutron irradiation. They are equivalent to small perfect dislocation loops and, in isolation in pure metals, undergo fast thermally-activated glide in the direction of their Burgers vector. Their strain field and ability to glide allows long-range interaction with other extended defects. Indeed, dislocations decorated by dislocation loops are commonly observed after neutron irradiation. Dislocations gliding under applied stress also encounter these mobile defects. These effects influence mechanical properties and require further investigation. This paper presents results from an atomic-scale study of copper and α-iron at either 0 K or 300 K. Loop drag and breakaway effects are investigated for an edge dislocation under applied stress interacting with a row of SIA loops below its glide plane. The maximum speed at which a loop is dragged is lower in co...

Copper implanted with helium at elevated temperatures has been examined by TEM. Helium bubbles with sizes larger than the size of bubbles in the interior of the grains had precipitated at some of the boundaries. The size of the bubbles in the boundaries tended to increase with increasing boundary energy.

An overview is given of the wide range of activities contained within recent R&D being performed within Europe on In-vessel materials for structural, heat sink and plasma facing purposes for ITER. The effect of creep-fatigue interaction on the fatigue life of CuCrZr and the effects of irradiation on over-aged CuCrZr are given. In addition the lifetime of ITER components has been further investigated by the performance of in situ experiments with both neutrons and high-energy protons. The effect of hydrogen on the crack initiation fracture toughness of Ti is reported at a range of irradiation temperatures. The irradiation induced stress relaxation of Alloy 718 used for bolting applications is being studied and initial results will be described. Work in the area of plasma facing materials and re-welding issues will also be presented.

Two copper alloys, dispersion-strengthened CuAl25 and precipitation-hardened CuCrZr, were examined under fatigue and fatigue with hold time loading conditions. Tests were carried out at room temperature and hold times were imposed at maximum tensile and maximum compressive strains. It was found that hold times could be damaging even at room temperature, well below temperatures typically associated with creep. Hold times resulted in shorter fatigue lives in the high cycle fatigue, long life regime (i.e., at low strain amplitudes) than those of materials tested under the same conditions without hold times. The influence of hold times on fatigue life in the low cycle fatigue, short life regime (i.e., at high strain amplitudes) was minimal. When hold time effects were observed, fatigue lives were reduced with hold times as short as two seconds. Appreciable stress relaxation was observed during the hold period at all applied strain levels in both tension and compression. In all cases, stresses relaxed quickly within the first few seconds of the hold period and much more gradually thereafter. The CuAl25 alloy showed a larger effect of hold time on reduction of high cycle fatigue life than did the CuCrZr alloy.

Over the last decade, molecular dynamics simulations of displacement cascades have revealed that glissile clusters of self-interstitial crowdions are formed directly in cascades and that they migrate one-dimensionally along close-packed directions with extremely low activation energies. Occasionally, under various conditions, a crowdion cluster can change its Burgers vector and glide along a different close-packed direction. The recently developed production bias model (PBM) of microstructure evolution under irradiation has been structured specifically to take into account the unique properties of the vacancy and interstitial clusters produced in the cascades. Atomic-scale kinetic Monte Carlo (KMC) simulations have played a useful role in understanding the defect reaction kinetics of one-dimensionally migrating crowdion clusters as a function of the frequency of direction changes. This has made it possible to incorporate the migration properties of crowdion clusters and changes in reaction kinetics into t...

Recently, annealed specimens of pure copper have been tensile tested in a fission reactor at a damage rate of 6 × 10−8 dpa/s with a constant strain rate of 1.3 × 10−7 s−1. The specimen temperature during the test was about 90 °C. The stress response was continuously recorded as a function of irradiation time (i.e. displacement dose and strain). The experiment lasted for 308 h. During the dynamic in-reactor test, the specimen deformed and hardened homogeneously without showing any sign of yield drop and plastic instability. However, the specimen yielded a uniform elongation of only about 12%. The preliminary results are briefly described and discussed in the present note.

In order to investigate the difference in defect accumulation between fcc Cu and bcc Fe, tensile specimens were neutron irradiated at ≃70 °C in the HFIR reactor at Oak Ridge National Laboratory to fluences in the range of 4.5×1020–4.7×1024 n/m2 (E>1 MeV) corresponding to displacement dose levels in the range of about 0.0001–0.8 dpa. Irradiated specimens were characterized using positron annihilation spectroscopy, transmission electron microscopy and electrical conductivity measurements. A limited number of iron specimens were also tensile tested. At 0.0001 dpa, a low density of very small vacancy clusters (1–3 vacancies) were detected in iron, while bigger three-dimensional cavities were observed at higher doses. Both their density and average size increased with increasing dose level. In contrast, no such cavities were observed in copper. Irradiation led to an increase in yield stress and a decrease in the uniform elongation for iron.

Two types of copper alloys, precipitation hardened (PH) Cu (CuCrZr-IG) and dispersion strengthened (DS) Cu (CuAl25-IG), are proposed as heat sink materials for the high heat flux (HHF) components of ITER. However, copper alloys are not included in any national codes, and properties of both CuCrZr and CuAl25 are not yet fully characterised. The performed R&D gives a basis for the specification of physical and mechanical properties required for the design analysis in accordance with the ITER Structural Design Criteria for In-vessel Components (SDC-IC). For both CuCrZr-IG and CuAl25-IG alloys, the statistical evaluation of available experimental data has been used to calculate the temperature dependence of the average value and of the 95% confidence limit of tensile properties. The stress limits, S m , S e , and S d , have been estimated on the basis of available data. The procedure used for specification of the temporary design limits of Cu alloys is described in the paper.

A series of kinetic Monte Carlo computer experiments performed on idealized systems clearly reveals the dramatic effects of 1-D migration of self-interstitial atom (SIA) crowdion clusters on the stability of void lattices. In the presence of migrating SIA, void lattices are shown to be unstable under pure 3-D SIA migration, but they are extremely stable, relative to random arrays of voids, under 1-D SIA migration. Void lattices remain stable even under the condition of fairly frequent changes in the Burgers vectors of the 1-D migrating SIA clusters. Clusters with average 1-D path segments having lengths on the order of the nearest neighbor distance in the void lattice can maintain the stability of void lattices.

Recent theoretical calculations and atomistic computer simulations have shown that glissile clusters of self-interstitial atoms (SIAs) play an important role in the evolution of microstructure in metals and alloys under cascade damage conditions. Over the past decade or so, the properties of SIA clusters in fcc, bcc and hcp lattices have been widely studied. In this paper we review key properties of these defects and also those of vacancy clusters formed directly in cascades, and present an atomic-level picture based on computer modelling of how these properties may change in the presence of other defects, impurities, stress fields, etc. We then examine the role of cluster properties and the consequences of their interactions in the process of damage accumulation and changes in mechanical and physical properties. We focus on the formation of defect clusters (e.g. dislocation loops and stacking fault tetrahedra (SFT)) and their segregation in the form of rafts of dislocation loops and atmospheres of loops decorating dislocations. Finally, we address the problem of radiation hardening by considering interactions between mobile dislocations and defect clusters (e.g. SIA dislocation loops, SFT and microvoids) produced during irradiation.

The analysis of the available experimental observations shows that the occurrence of a sudden yield drop and the associated plastic flow localization are the major concerns regarding the performance and lifetime of materials exposed to fission or fusion neutrons. In the light of the known mechanical properties and microstructures of the as-irradiated and irradiated and deformed materials, it has been argued that the increase in the upper yield stress, the sudden yield drop and the initiation of plastic flow localization, can be rationalized in terms of the cascade induced source hardening (CISH) model. Various aspects of the model (main assumptions and predictions) have been investigated using analytical calculations, 3-D dislocation dynamics and molecular dynamics simulations. The main results and conclusions are briefly summarized. Finally, it is pointed out that even though the formation of cleared channels may be rationalized in terms of climb-controlled glide of the source dislocation, a number of problems regarding the initiation and the evolution of these channels remain unsolved.

A large number of displacement cascades has been simulated using molecular dynamics (MD) for primary knock-on atoms (PKAs) of energy in the range 2–20 keV at 100 and 600 K. The defect production efficiency, intracascade clustering efficiency and fractions of self-interstitial atoms (SIAs) and vacancies contained in the clusters produced in the cascades were determined. The fractions of both SIAs and vacancies contained in clusters increased with increasing PKA energy and approached a saturation level at 20 keV. At both temperatures the clustered fraction of SIAs was higher than the clustered fraction of vacancies. This difference is larger at 600 K than that at 100 K and increases with increasing PKA energy. These results are compared with the corresponding values derived from experimental results. The comparison suggests that the single cascade simulations yield considerably higher values of the clustered fractions of both SIAs and vacancies primarily because the impact of intercascade defect reaction kinetics cannot be taken into account in the MD simulations.

In the unirradiated condition the Ti6Al4V (α+β) alloy has slightly higher tensile strength and noticeably lower ductility compared to that of the Ti5Al2.5Sn (α) alloy both at 50 and 350 °C. The fracture toughness behaviour of both alloys is similar at ambient temperature. At 350 °C, on the other hand, the fracture toughness of the (α) alloy is lower compared to that of the (α+β) alloy. Neutron irradiation at 50 °C to a dose level of 0.3 dpa caused hardening, plastic instability and a substantial reduction in fracture toughness of both alloys. Irradiation at 350 °C resulted in a substantial hardening and a significant decrease in the fracture toughness in the (α+β) alloy due to irradiation induced precipitation whereas only minor changes in the tensile and fracture toughness behaviour were observed in the (α) alloy. The tensile and fracture toughness properties of the (α+β) alloy are more strongly affected by neutron irradiation compared to that of the (α) alloy.

The tensile and fracture behaviour of CuCrZr and CuAl25 IG0 alloys joint to 316L(N) stainless steel by hot isostatic pressing (HIP) have been determined in unirradiated and neutron-irradiated conditions. The tensile and fracture behaviour of copper alloy HIP joint specimens are dominated by the properties of the copper alloys, and particularly, by the strength mismatch and mismatch in strain hardening capacities between copper alloys and stainless steel. The test temperature, neutron irradiation and thermal cycles primarily affect the copper alloy HIP joint properties through changing the strength mismatch between the base alloys. Changes in the loading conditions i.e. tensile, bend ( J I ) and mixed-mode bend ( J I / J II ) lead to different fracture modes in the copper alloy HIP joint specimens.

The effect of post-irradiation annealing at 300 °C for 50 h on the microstructure and mechanical properties of CuCrZr irradiated to 0.2 and 0.3 dpa at 100 °C has been evaluated. The post-irradiation annealing removes the yield point phenomenon, lowers the strength and improves the ductility and work hardening of the material. It does not, however, completely remove the effects of irradiation as the annealed material is still considerably stronger and less ductile than the unirradiated CuCrZr alloy. The removal of the yield point phenomenon is attributed to a decrease in raft formation and dislocation decoration, weakening the dislocation pinning and allowing more homogeneous generation of dislocations within the material. The segregation of solute additions, impurities and transmutation elements as well as precipitation at dislocation segments and loops needs to be further investigated to fully understand the mechanical properties.

Specimens of oxygen-free high conductivity (OFHC) copper were irradiated in the DR-3 reactor at Riso at 100 °C to doses in the range 0.01–0.3 dpa (NRT). Some of the specimens were tensile tested in the as-irradiated condition at 100 °C whereas others were given a post-irradiation annealing treatment at 300 °C for 50 h and subsequently tested at 100 °C. The microstructure of specimens was characterized in the as-irradiated as well as irradiated and annealed conditions both before and after tensile deformation. While the interstitial loop microstructure coarsens with irradiation dose, no significant changes were observed in the population of stacking fault tetrahedra (SFT). The post-irradiation annealing leads to only a partial recovery and the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade-induced source hardening (CISH) model.

Over the years, an enormous amount of experimental results have been reported on damage accumulation (e.g. void swelling) in metals and alloys irradiated under vastly different recoil energy conditions. Unfortunately, however, very little is known either experimentally or theoretically about the effect of recoil energy on damage accumulation. Recently, dedicated irradiation experiments using 2.5 MeV electrons, 3.0 MeV protons and fission neutrons have been carried out to determine the effect of recoil energy on the damage accumulation behaviour in pure copper and the results have been reported in part I of this paper (Singh et al., 2001, Phil. Mag. A, 80. 2629). The present paper attempts to provide a theoretical framework within which the effect of recoil energy on damage accumulation behaviour can be understood. The damage accumulation under Frenkel pair production (e.g. 2.5 MeV electron) has been treated in terms of the standard rate theory (SRT) model whereas the evolution of the defect micro...

Screening experiments were carried out to determine the effect of bonding and bakeout thermal cycles on microstructure, mechanical properties and electrical resistivity of the oxide dispersion strengthened (GlidCop, CuAl-25) and the precipitation hardened (CuCrZr, CuNiBe) copper alloys. Tensile specimens were given heat treatments corresponding to solution anneal, prime-aging, bonding thermal treatment followed by re-aging and the reactor bakeout treatment. A number of specimens were irradiated at 350°C to a dose level of ≃0.3 dpa in the DR-3 reactor at Riso. Both unirradiated and irradiated specimens were tensile tested at 350°C. The microstructure and electrical resistivity were determined in the unirradiated and irradiated conditions. The post-deformation microstructure and fracture surfaces of unirradiated and irradiated specimens were examined. The main results are described and their salient features discussed. The most significant effect of neutron irradiation is a severe loss of ductility in the case of CuNiBe alloys.

An approximate method for the numerical solution of a kinetic equation describing the nucleation, growth and coarsening of point-defect clusters or second-phase precipitates is of interest for many applications in solid-state physics. In the present paper the validity of a grouping method developed by Kiritani (1973, J. phys. Soc. Japan 35, 95) is examined. In this method, which is the only one proposed up to now, a group of equations is replaced by an ‘averaged’ equation. Significant disagreement between the group kinetic equation and the original kinetic equation is revealed and the reasons for this are specified. A new grouping method based on the original equation is proposed and application of the method for the solution of two problems is demonstrated by comparing the results of numerical and analytical calculations.

Collision cascades in metals produce small interstitial clusters and perfect dislocation loops that glide in thermally activated one-dimensional (1D) random walks. These gliding defects can change their Burgers vectors by thermal activation or by interactions with other defects. Their migration is therefore `mixed 1D/3D migration' along a 3D path consisting of 1D segments. The defect reaction kinetics under mixed 1D/3D diffusion are different from pure 1D diffusion and pure 3D diffusion, both of which can be formulated within analytical rate theory models of microstructure evolution under irradiation. Atomic-scale kinetic Monte Carlo (kMC) defect migration simulations are used to investigate the effects of mixed 1D/3D migration on defect reaction kinetics as a guide for implementing mixed 1D/3D migration into the analytical rate theory. The functional dependence of the sink strength on the size and concentration of sinks under mixed 1D/3D migration is shown to lie between that for pure 1D and pure 3D migration and varies with L, the average distance between direction changes of the gliding defects. It is shown that the sink strength in simulations for spherical sinks of radius R under mixed 1D/3D migration for values of L greater than R can be approximated by an expression that varies directly as R2. For small L, the form of the transition from mixed 1D/3D to pure 3D diffusion as L decreases is demonstrated in the simulations, the results of which can be used in the future development of an analytical expression describing this transition region.

Transmission electron microscope (TEM) disks of pure copper and copper containing 0.17–10% Ni were neutron irradiated at 210°C and 420°C in Hf-shielded capsules in the High Flux Isotopes Reactor to doses of 13.5 and 14.9 displacements per atom (dpa), respectively. Void swelling was not observed in any of the specimens irradiated at 210°C. Instead, a high density of stacking fault tetrahedra (SFTs) and a moderate density of dislocation loops were observed. There was no evidence for defect cluster patterning (wall formation) in any of the specimens irradiated at 210°C. The SFT density was independent of Ni content, whereas the loop density was highest in the alloy containing ∼2% Ni. Pronounced void swelling was observed in all of the specimens irradiated at 420°C. A void denuded zone of ∼2 μm width was observed adjacent to grain boundaries. The void swelling in copper containing ∼2% Ni showed a pronounced maximum in a 10-μm wide band adjacent to the grain boundary denuded zones. Matrix voids aligned along 〈1 1 0〉 directions were observed in Cu–10%Ni irradiated at 420°C. The occurrence of the peak swelling zone and the partial ordering of voids are both consistent with the predictions of the recent production bias model which includes 1-D diffusional transport of clusters of self-interstitial atoms.

Recent theoretical calculations and atomistic computer simulations have shown that one-dimensional glissile clusters of self-interstitial atoms (SIAs) play an important role in the evolution of microstructure in metals and alloys under cascade damage conditions. Recently, it has been proposed that the evolution of heterogeneities such as dislocation decoration and rafts has serious impacts on the mechanical properties on neutron-irradiated metals. In the present work, atomic-scale computer modelling (ASCM) has been applied to study the mechanisms for the formation of such microstructure in bcc iron. It is shown that glissile clusters with parallel Burgers vectors interact strongly and can form extended immobile complexes, i.e., rafts. Similar attractive interaction exists between dislocation loops and an edge dislocation. These two mechanisms may be responsible for the formation of extended complexes of dislocation loops below the extra half-plane of edge dislocations. The interaction energies between loops and between an edge dislocation and loops has been calculated as a function of distance using ASCM and the results for long-range interactions are in good agreement with the results of isotropic elasticity calculations.

In recent years, it has been shown that intra-cascade clustering of vacancies and self-interstitial atoms (SIAs), differences in the thermal stability and mobility of the resulting clusters and one-dimensional (1-D) diffusional glide of SIA clusters play a key role in damage accumulation in metals under cascade damage conditions. The model taking these aspects into account (production bias model, PBM) succeeded in rationalising striking features in the microstructural evolution in pure metals, where the conventional rate theory model failed: the high overall swelling even at low dislocation densities, the enhanced swelling near grain boundaries, the decoration of dislocations with SIA loops, saturation of void growth and void lattice formation. In the present paper, the main ideas and results of these considerations are reviewed. We discuss recent work on possible effects of deviations of SIA cluster diffusion from strictly 1-D by direction changes and/or self-climb and formulate a general reaction kinetics including 1-D and 3-D cluster diffusion. Such reaction kinetics may be considered to form the basis for a general description of cascade damage accumulation in metals and complex technical alloys.

Static and molecular dynamics simulations have been used with different types of interatomic potentials to investigate the structure, properties and stability of self-interstitial atom (SIA) clusters produced during irradiation. In α-iron (Fe), faulted clusters of dumbbells are unstable for all the potentials. The most stable SIA clusters are sets of parallel crowdions. Large clusters of this type form perfect dislocation loops with Burgers vector b = ½〈111〉. Small clusters (less than 9 SIAs) of 〈100〉 crowdions are stable at 0K, but transform into a set of 〈111〉 crowdions on annealing. Larger 〈100〉 clusters are stable and form perfect dislocation loops with b = 〈100〉. Both types of loops are glissile. In copper (Cu), clusters of parallel 〈100〉 dumbbells and 〈110〉 crowdions are stable. Large clusters of these types form faulted and perfect dislocation loops with b = ⅓ 〈111〉 and ½ 〈110〉 respectively. Small faulted clusters (less than 7 SIAs) of irregular shape can transform into a set o...

To compare the defect accumulation and the annealing behaviour in an fcc and a bcc metal, OFHC-Cu and pure Fe were neutron irradiated at 100°C to a fluence of 1.5 × 1024 n/m2, (E > 1 MeV). Isochronal annealing was carried out and the annealing behaviour followed by positron annihilation spectroscopy (PAS) as well as electrical conductivity measurements. The results for the two specimens in the as-irradiated state are very different. In Cu the defect positron lifetime is characteristic of single vacancies, very small vacancy clusters or stacking fault tetrahedra, while in Fe the defect lifetimes confirm the presence of micro-voids and voids. The electrical conductivity, on the other hand does not discriminate between the two types of damage in the irradiated specimens. During annealing of the irradiated Fe below stage V, the average void size grows by migration and coalescence of the micro-voids and voids. At and above stage V the void density decreases and the voids finally anneal out at ∼500°C. In contrast, the annealing of irradiated Cu below stage V does not yield any evidence for the evolution of micro-voids or voids. The implications of these results are discussed. One conclusion is that neutron irradiation below stage V causes higher void swelling in bcc iron than in fcc copper.

According to the production bias model, glissile defect clusters and small dislocation loops play an important role in the microstructural evolution during irradiation under cascade damage conditions. The atomic scale computer simulations carried out in recent years have clarified many questions about the structure and properties of glissile clusters of self-interstitial atoms that are formed directly in the cascade volume. It has been found that such clusters consist of sets of crowdions and are highly mobile in the crowdion direction. Very recently, one-dimensional glide of similar character has been observed in the computer simulation of small vacancy loops in α-Fe. In the present paper we summarise results obtained by molecular dynamics simulations of defect clusters and small dislocation loops in α-Fe(bcc) and Cu(fcc). The structure and stability of vacancy and interstitial loops are reviewed, and the dynamics of glissile clusters assessed. The relevance and importance of these results in establishing a better understanding of the observed differences in the damage accumulation behaviour between bcc and fcc metals irradiated under cascade damage conditions are pointed out.

Accumulation of nano-size prismatic defect clusters near slip-dislocations results from their mutual elastic interaction. We present here 3-D isotropic elasticity calculations for the interaction energy between radiation-induced nano-size prismatic loops and grown-in dislocation loops. The current treatment extends the work of Trinkaus et al. in two respects. First, a computational method for full 3-D analysis of interaction energies in bcc Fe and fcc Cu is developed. Second, the theoretical method of Kroupa is computationally implemented for rigorous calculations of force, torque and induced surface energy on defect clusters. It is shown that small clusters are trapped within a zone of ∼10 nm in bcc Fe, and ∼20 nm in fcc Cu at room temperature, in rough agreement with experimental observations. Clusters can be absorbed in the core of grown-in dislocations because of unbalanced moments, which provide sufficient energy for rotation of their Burgers vectors in a zone of 2–3 nm in Fe. Near the dislocation core (within a few nanometers), sessile defect clusters in Cu are shown to convert to a glissile configuration.

Kinetic Monte Carlo simulations of one-dimensionally diffusing interstitial clusters (dislocation loops) are used to gain insight into their role in microstructure evolution under irradiation. The simulations investigate the changes in reaction kinetics of these defects as a function of changes in the Burgers vector and variation in the size and density of randomly or periodically distributed sinks. In this paper we report on several kinetic Monte Carlo studies intended to elucidate the effects of mixed 1-D/3-D migration relative to pure 3-D and pure 1-D migration. We have investigated the effects of variation of the average distance traveled between Burgers vector changes 〈L〉 on the absorption of individual defects into absorbers of varying size and varying concentration, as well as the effects of variation in 〈L〉 on the time dependence of absorption of a collection of defects into an array of absorbers. Significant effects of Burgers vector changes on the reaction kinetics of the diffusing interstitial clusters are clearly demonstrated. Even when 〈L〉 is large relative to the size and spacing of microstructural features, significant effects of mixed 1-D/3-D migration on reaction kinetics are evident.

Crack nucleation and growth behaviour are important parameters in deciding about the applicability of the dispersion strengthened copper alloy CuAl25 in components such as the first wall and divertor in ITER. The effective strain to fracture of notched tensile specimens decreased with increasing stress state triaxiality and with increasing temperature at constant constraint level following the Rice and Tracey model for void growth. In three point bend tests, the strain for stable crack initiation decreased significantly with increasing temperature. The CuAl25 alloy failed by a ductile microvoid mechanism where extensive void nucleation occurred at very low strains at grain boundaries with increasing stress state triaxiality. At elevated temperatures the fracture surface morphology changed from microvoid to intergranular fracture in three-point bend tests.

The fatigue performance of pure copper of the oxygen free, high conductivity (OFHC) grade and a dispersion strengthened copper alloy, Glid Cop TM CuAl-25 was examined with and without irradiation exposure. Mechanical testing was carried out to establish the fatigue lives of these materials in the unirradiated and irradiated states. Fatigue specimens of these two materials were irradiated with fission neutrons in the DR-3 reactor at Riso with a flux of ≈2.5 × 10 −17 n/m 2 s ( E > 1 MeV) to fluence levels of 1.5–2.5 × 10 24 n/m 2 ( E > 1 MeV) at ≈47 and 100°C. Specimens irradiated at 47°C were fatigue tested at room temperature, whereas those irradiated at 100°C were tested at the irradiation temperature. These investigations demonstrated that, while irradiation causes significant hardening of the materials, the hardening appears to have no negative impact on fatigue performance. In fact, the fatigue performance of the CuAl-25 alloy is considerably better in the irradiated than that in the unirradiated state tested both at 22°C and 100°C. Microstructural observations of the fatigue microstructures and the fracture surfaces are useful in understanding this conclusion.

Tensile specimens of pure iron were irradiated with fission neutrons (i) at 320 K to displacement dose levels of 7.5 × 10−3, 7.5 × 10−2 and 3.75 × 10−1 dpa (NRT) and (ii) at 523 K to dose levels of 7.5 × 10−2 and 2.25 × 10−1 dpa (NRT). Both unirradiated and irradiated specimens were tensile tested at the irradiation temperatures. Microstructures of the as-irradiated and irradiated and tensile tested specimens were investigated by transmission electron microscopy. Fracture surfaces of tensile tested specimens in unirradiated and irradiated conditions were examined in a scanning electron microscope. Results of these investigations are reported and discussed particularly in terms of the role of interstitial clusters (produced directly in the cascades) and their transport via one-dimensional glide. It is suggested that the formation and interaction of `cleared' channels may play a significant role in determining the deformation and fracture behaviour of the irradiated pure iron.

Defect clusters form readily in collision cascades in metals, and some of the self-interstitial atom clusters form as crowdion clusters that diffuse by one-dimensional migration along a close-packed direction. Defect interactions and thermal fluctuations can cause the direction of one-dimensional migration to change, resulting in a mixed one-dimensional/ three-dimensional migration. Kinetic Monte Carlo computer simulations applied to model systems are used to investigate the effects of one-dimensional, three-dimensional and mixed one-dimensional/ three-dimensional migration on defect reaction kinetics. The functional relationships between the sink strength, the size of sinks and the average distance between direction changes during mixed one-dimensional/three-dimensional migration are explored.

Considerations of intracascade clustering of self-interstitial atoms (SIAs) and vacancies during the cooling-down phase of a cascade and of the difference between thermal stabilities of SIA and vacancy clusters give rise to the concept of production bias. In recent years, various aspects of this concept including the effects of cluster annihilation by one-dimensional glide have been investigated. The salient features of these investigations and their results are summarized. The most important conclusion emerging from these investigations is that a realistic modelling of accumulation of surviving defects under cascade damage conditions must explicitly consider the role of formation, thermal stability and mobility (glide and climb) of clusters produced in the cascades. Examples are shown where appropriate considerations of these features have led to predictions regarding dose, temperature and recoil energy dependence of void swelling which are consistent with experimental results. The special featu...

In metals and alloys subjected to cascade damage dislocations are frequently found to be decorated with a high density of small clusters of self-interstitial atoms (SIAs) in the form of dislocation loops. In the present paper it is shown that this effect may be mayributed to the glide and trapping of SIA loops, produced directly in cascades (rather than to the enhanced agglomeration aggsingle SIAs), in the strain field of the dislocations. The conditions for the accumulation of glissile SIA loops loor dislocations as well as the dose and temperature dependencies of this phenomenon are discussed. It is suggested that the decoration of dislocations with loops may play a key role in radiation hardening subjected to cascade damage. It is shown, for example, that the increase in the upper yield stress followed by a yield drop and plastic instability in metals and alloys subjected to cascade damage cannot be rationalized in terms of conventional dispersed barrier hardening (DBH) but may be understood in terms of cascade induced source hardening WISH) in which the dislocations are considered to be locked by the loops decorating them. Estimates for the stress necessary to pull a dislocation away from its loop ‘cloud’ are used to discuss the dose and temperature dependence of plastic flow initiation.

The conventional theoretical models describing the damage accumulation, particularly void swelling, under cascade damage conditions do not include treatments of important features such as intracascade clustering of self-interstitial atoms (SIAs) and one-dimensional glide of SIA clusters produced in the cascades. Recently, it has been suggested that the problem can be treated in terms of ‘production bias’ and one-dimensional glide of small SIA clusters. In the earlier treatments a ‘mean size approximation’ was used for the defect clusters and cavities evolving during irradiation. In the present work, we use the ‘size distribution function’ to determine the dose dependence of sink strengths, vacancy supersaturation and void swelling as a function of dislocation density and grain size within the framework of production bias model and glide of small SIA clusters. In this work, the role of the sessile-glissile loop transformation (due to vacancy supersaturation) on the damage accumulation behaviour is included. The calculated results on void swelling are compared with the experimental results as well as the results of the earlier calculations using the ‘mean size approximation’. The calculated results agree very well with the experimental results.

A brief introduction to positron annihilation spectroscopy (PAS), and in particular to its use for defect studies in metals is given. Positrons injected into a metal may become trapped in defects such as vacancies, vacancy clusters, voids, bubbles and dislocations and subsequently annihilate from the trapped state in the defect. The annihilation characteristics (e.g., the lifetime of the positron) can be measured and provide information about the nature of the defect (e.g., size, density, morphology). The technique is sensitive to both defect size (in the range from monovacancies up to cavities containing 50–100 vacancies) and density in metals. Monovacancies can typically be detected in concentrations higher than a tenth of a part per million. For three dimensional vacancy clusters the sensitivity increases with increasing cluster size. The combination of PAS with theoretical calculations and with other experimental methods (in particular transmission electron microscopy) forms the basis for the use of PAS to quantitatively characterize defects and defect complexes, both visible and invisible in transmission electron microscopes; this is illustrated by some examples. Finally, the advantages of the use of PAS are pointed out.

Tensile specimens of CuAl 2 O 3 , CuCrZr and CuNiBe alloys were given different heat treatments and then irradiated with fission neutrons at 250°C to a dose level of ≅ 0.3 dpa. Both unirradiated and irradiated specimens were tensile tested at 250°C. The microstructure and electrical resistivity were determined in the unirradiated as well as irradiated conditions. The post-deformation microstructure and fracture surfaces were also investigated. The main effect of the bonding thermal cycle heat treatment was a slight decreased in the strength of CuCrZr and CuNiBe alloys. The strength of CuAl-25, on the other hand, remained almost unaltered. The post-irradiation tests at 250°C showed a severe loss of ductility in the case of CuNiBe alloy. The irradiated CuAl-25 and CuCrZr specimens, on the other hand, exhibited a reasonable amount of uniform elongation. The results are briefly discussed in terms of thermal and irradiation stability of precipitates and particles and irradiation-induced segregation, precipitation and recovery of dislocation microstructure.

The deformation microstructures of oxygen-free high-conductivity (OFHC) copper were examined by transmission electron microscopy (TEM) following in-reactor and post-irradiation slow strain rate tensile tests. The TEM results suggest that the main modes of deformation differ between all examined cases. Plastic deformation appeared predominantly localized in the defect-free cleared channels following post-irradiation testing and hardly any dislocations were seen outside the channels. The microstructures following in-reactor tests were characterized by a small amount of cleared channels and a distinct dislocation density within the matrix. However, the dislocations observed following in-reactor testing did not seem to interact with each other, whereas that was the main mode of deformation in the non-irradiated reference sample. The possible mechanisms of plastic deformation are discussed based on the experimental results.

The electrical resistivity and tensile properties of copper and oxide dispersion strengthened (DS) copper alloys have been measured before and after fission neutron irradiation to damage levels of 0.5 to 5 displacements per atom (dps) at {approximately}100 to 400{degrees}C. Some of the specimens were irradiated inside a 1.5 mm Cd shroud in order to reduce the thermal neutron flux. The electrical resistivity data could be separated into two components, a solid transmutation component {Delta}{rho}{sub tr} which was proportional to thermal neutron fluence and a radiation defect component {Delta}{rho}{sub rd} which was independent of the displacement dose. The saturation value for {Delta}{rho}{sub rd} was {approximately}1.2 nanohm-meters for pure copper and {approximately}1.6 nanohm-meters for the DS copper alloys irradiated at 100{degrees}C in positions with a fast-to-thermal neutron flux ratio of 5. Considerable radiation hardening was observed in all specimens at irradiation temperatures below 200{degrees}C. The yield strength was relatively insensitive to neutron spectrum in specimens strengthened by dispersoids or cold- working. 17 refs., 7 figs., 1 tab.

It is well established that clusters of both vacancies and self-interstitial atoms (SIAs) are produced in displacement cascades. Ample evidence has also been presented showing that SIAs produced in the form of small dislocation loops may be highly glissile. Such loops may glide to and may be absorbed by extended sinks such as dislocations and grain boundaries (GBs). The loss of SIAs by this process causes a vacancy supersaturation representing an efficient driving force for void swelling, in particular in regions adjacent to GBs. Enhanced swelling in regions adjacent to GBs has been observed in several metals subject to irradiation by both fast fission neutrons and 600 MeV protons. In the latter case, however, the width of the region of enhanced swelling is smaller and the amount of swelling is significantly lower than in the former case. Recently, enhanced swelling near GBs as induced by the cascade damage of fast neutrons has been discussed in terms of SIA loop escape to GBs assuming that the range of glissile loops and, via this, the width of the peak zone, is controlled by the visible void structure. In the present paper, this model is applied to irradiation with 600 MeV protons where the cascade damage is accompanied by a high helium production rate. It is shown that, in this case, the width of the peak zone is controlled by the (mostly invisible) bubble structure rather than by the (visible) void structure. The reduced swelling relative to that under neutron irradiation is attributed to the screening of the GBs with respect to loop capture by the bubble structure in the void denuded zone.