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U3Si2 is an advanced fuel candidate due to its relatively high fissile density and attractive thermal properties. Compared to standard UO2 fuel, there are significant data gaps for the thermophysical and thermomechanical properties of U3Si2. Point defect concentrations and mobilities under irradiation govern a number of important fuel performance properties, such as creep and fission gas release. In this work, we utilized density functional theory (DFT) data to inform a cluster dynamics framework to predict point defect concentrations in U3Si2 under irradiation. Molecular dynamics (MD) simulations were used to examine the contribution of atomic mixing during ballistic cascades to diffusion, as well as the diffusivity of U and Si at grain boundaries. These atomic scale models for diffusivity were then used to inform a creep model based on bulk (Nabarro-Herring) and grain boundary (Coble) diffusional creep, and climb-limited dislocation creep. The model compares well against available experimental data and has been implemented in the BISON fuel performance code. A demonstration case using simple power profiles has been carried out, showing that negligible creep occurs due to the low temperatures experienced by U3Si2 in-reactor, a consequence of its high thermal conductivity.

In this paper, a combination of quantum mechanical and interatomic potential-based atomistic calculations are used to predict the core structures of screw and edge dislocations in tungsten in the presence of a particular concentration of hydrogen atoms. These configurations of the core structures are the results of two competing energies: the interaction between the partial dislocations and the corresponding generalized stacking fault energy in between the two partial dislocations, which are presented in this work. With this, we can precisely predict the configurations of the hydrogen-doped dislocation core structures.

Out-of-reactor, non-irradiated thermal creep performances and lives of annealed and stress-relieved Zr-1.02Nb-0.69Sn-0.12Fe cladding tubes were studied and compared. The creep rates of annealed Zr-1.02Nb-0.69Sn-0.12Fe cladding tubes were appreciably slower than those of stress-relieved annealed counterpart. The stress exponent increased slightly from 5.1 to 6.1 in the stress-relieved cladding to 5.3–6.3 in the annealed cladding. The creep activation energy of the annealed Zr-1.02Nb-0.69Sn-0.12Fe alloy (300–330 kJ/mol) was larger compared to that of the stress-relieved alloy (210–260 kJ/mol). The creep activation energy of annealed alloy is close to that of self-diffusion in α-Zr (336 kJ/mol). The smaller activation energy in the stress-relieved alloy is attributed to the increasing contribution of faster diffusion path such as grain boundaries and dislocations. The presence of dislocation arrays with higher dislocation density and smaller grain size in the stress-relived alloy was confirmed by TEM analysis. The creep rupture time increased dramatically in the annealed Zr–1Nb- 0.7Sn-0.1Fe alloy compared to that of stress-relieved alloy, supporting the decrease of creep rate by annealing. The creep life of Zr-1.02Nb-0.69Sn-0.12Fe claddings can be extended through microstructure modification by annealing at intermediate temperatures in which dislocation creep dominates.

An application of high-energy wide angle synchrotron X-ray diffraction to investigate the tensile deformation of 9Cr ferritic/martensitic (F/M) ODS steel is presented. With tensile loading and in-situ X-ray exposure, the lattice strain development of matrix was determined. The lattice strain was found to decrease with increasing temperature, and the difference in Young's modulus of six different reflections at different temperatures reveals the temperature dependence of elastic anisotropy. The mean internal stress was calculated and compared with the applied stress, showing that the strengthening factor increased with increasing temperature, indicating that the oxide nanoparticles have a good strengthening impact at high temperature. The dislocation density and character were also measured during tensile deformation. The dislocation density decreased with increasing of temperature due to the greater mobility of dislocation at high temperature. The dislocation character was determined by best-fit methods for different dislocation average contrasts with various levels of uncertainty. The results shows edge type dislocations dominate the plastic strain at room temperature (RT) and 300 °C, while the screw type dislocations dominate at 600 °C. The dominance of edge character in 9Cr F/M ODS steels at RT and 300 °C is likely due to the pinning effect of nanoparticles for higher mobile edge dislocations when compared with screw dislocations, while the stronger screw type of dislocation structure at 600 °C may be explained by the activated cross slip of screw segments.

We present here an application of a fundamentally new theoretical framework for description of the simultaneous evolution of radiation damage and plasticity that can describe both in situ and ex situ deformation of structural materials [1]. The theory is based on the variational principle of maximum entropy production rate; with constraints on dislocation climb motion that are imposed by point defect fluxes as a result of irradiation. The developed theory is implemented in a new computational code that facilitates the simulation of irradiated and unirradiated materials alike in a consistent fashion [2]. Discrete Dislocation Dynamics (DDD) computer simulations are presented here for irradiated fcc metals that address the phenomenon of dislocation channel formation in post-irradiated copper. The focus of the simulations is on the role of micro-pillar boundaries and the statistics of dislocation pinning by stacking-fault tetrahedra (SFTs) on the onset of dislocation channel and incipient surface crack formation. The simulations show that the spatial heterogeneity in the distribution of SFTs naturally leads to localized plastic deformation and incipient surface fracture of micro-pillars.

Extensive plastic deformation in the metal underneath the oxide scale in autoclave tested Zircaloy-2 was studied using transmission electron microscopy (TEM). It was concluded that the plastic deformation is created by creep during oxidation, and is not caused by surface treatment, sample preparation or cooling from autoclave temperatures. Evidence of large strains was found in the form of dislocation tangles, dis- location patches and sub-grain formation, and also indications of twinning were found. The heavily deformed layer is around a few lm thick and no obvious difference could be seen between alloys with different strength or different oxide thickness.

Dislocation dynamics is used to investigate the distribution of dislocations in oxide nuclear fuel under irradiation using the values of dislocation density from experiments. A model is constructed to account for the effects of irradiation on dislocation movement and for the brittle behavior of the material. Results show that the ground state of interacting dislocations in UO 2 during irradiation is a periodic structure with spacing between walls equal to 100–300 nm at experimental dislocation densities. These regions adorned by dislocation walls are called sub-grains and represent the result of polygonization. The threshold of polygonization is shown to depend on the fluctuations of the stress field produced by interaction of many dislocations. These fluctuations reach a critical value when a critical dislocation density is reached (∼4 × 10 14 m −2 ). The calculated value matches experimental data on dislocation density measurement of irradiated uranium dioxide at burn-up corresponding to the formation of high burn-up structure.

The plasticity of the fluorite structure in UO 2 is investigated with molecular dynamics simulation and empirical potential. The stacking fault energies and the dislocation core structures with Burgers vector a 2 〈 1 1 0 〉 are systematically calculated. All dislocation core structures show a significant increase of the oxygen sub-lattice disorder at temperatures higher than 1500 K. The threshold stress for dislocation glide is found to decrease with increasing temperature but its values is always very high, several GPa at 0 K and several hundred of MPa at 2000 K. A relation between the dislocation mobility dependence with temperature and the increase of the oxygen sub-lattice disorder in the dislocation cores is established.

Dislocation evolution during thermal creep deformation was investigated at 1023 K for V–4Cr–4Ti alloys with various thermal and mechanical treatments. Changes in the density and Burgers vector of dislocations were examined with various materials and experimental variables (e.g., cold work introducing dislocations, thermal aging introducing high density of Ti-rich precipitates, applied stress, and creep deformation levels). The Burgers vector analysis showed that, in the annealed specimens, dislocations induced by the thermal creep were predominantly of a /2〈1 1 1〉 type. The densities of a 〈1 0 0〉 and a /2〈1 1 1〉 type dislocations were comparable for the cold worked V–4Cr–4Ti, but the fraction of a /2〈1 1 1〉 type dislocations increased with the creep deformation. Similar changes to the Burgers vector fraction were observed for specimens with a high density of precipitation prior to the cold work. On the other hand, cold work followed by aging (i.e., strain aging) was effective in keeping dislocation structures during the creep deformation.

In situ transmission electron microscopy (TEM) observation of UO2 single crystal irradiated with Kr ions at high temperatures was conducted to understand the dislocation evolution due to high-energy radiation. The dislocation evolution in UO2 single crystal is shown to occur as nucleation and growth of dislocation loops at low-irradiation doses, followed by transformation to extended dislocation segments and networks at high doses, as well as shrinkage and annihilation of some loops and dislocations due to high temperature annealing. Generally the trends of dislocation evolution in UO2 were similar under Kr irradiation at different ion energies and temperatures (150 keV at 600 °C and 1 MeV at 800 °C) used in this work. Interstitial-type dislocation loops with Burgers vector along 〈1 1 0〉 were observed in the Kr-irradiated UO2. The irradiated specimens were denuded of dislocation loops near the surface.

Alloy 617 is considered a leading structural material for next generation nuclear power plants due to its good corrosion resistance and exceptional strength at high-temperatures. In the present work, a complementary study of deformation mechanisms in Alloy 617 was performed via tensile testing at temperatures up to 1000 °C. Dynamic strain aging characterized by serrated flow leading to temperature independent yield strength was observed at intermediate temperatures. At temperatures higher than 700 °C, the material’s strength reduced dramatically as a result of the predominance of dislocation creep and dynamic recrystallization. The changes in deformation mechanisms were established by direct observations of Electron Backscatter Diffraction mappings and by the analysis of texture development based on the pole figure mappings. Recrystallized grains were found to preferentially grow in particle-rich areas and on high-angle grain boundaries. Finally, fracture was initiated from particle cracks at temperatures up to 700 °C and by triple junction cracks from 800 to 1000 °C.

The interaction process between dislocations and large stacking-fault tetrahedra was observed in real time at high temperature during deformation experiments in situ in the transmission electron microscope. Dislocation interactions with tetrahedra resulted in them being annihilated and converted to another defect type. Dislocation bypass of the tetrahedra occurred by cross-slip. The latter interaction occurred slowly and halted the progress of the dislocation. Annihilation versus bypass by dislocation cross-slip was dictated by the location at which the slip plane intersected the tetrahedron – on the face or along the edges with the stair-rod dislocations. In general, the interactions, at best, were weakly temperature dependent. Published by Elsevier B.V.

The new model for the vacancy field evolution in grains during annealing of irradiated fuel was developed and implemented in the MFPR code. The model simulates time and spatial variation of the vacancy concentration in the presence of extended vacancy sources (grain boundaries and dislocations) and sinks (growing intragranular bubbles). Being combined with the models for dislocation creep and for bubbles biased migration in the vacancy gradient, the new model self-consistently describes the processes of gas release and microstructure evolution observed in the annealing tests.

Oxide dispersion strengthened (ODS) steels are promising candidates for applications in fusion and fission rectors. Y 2 O 3 particles dispersed in an iron matrix drastically improve the strength without adverse effects on ductility. We investigate here details of the dislocation core structure in Y 2 O 3 precipitates, and at the interface between the iron matrix and the Y 2 O 3 precipitate. We also simulate dislocation interaction with nano-size Y 2 O 3 precipitates. It is shown that the γ -surface energies on planes between oxygen atoms and metallic atoms are extremely high, and that dislocations in an iron matrix cannot penetrate into Y 2 O 3 particles. Three-dimensional parametric dislocation dynamics simulations of the interaction between an edge dislocation and an Y 2 O 3 particle are carried out. The results show that the critical resolved shear stress (CRSS) has a strong dependence on the lattice mismatch Y 2 O 3 particles and the iron matrix, and that it is lower than analytically calculated values of the Orowan stress.

Creep experiments performed at 650 °C and 250 MPa on MA957 and CEA-developed 18%Cr ferritic Oxide Dispersion Strengthened (ODS) steels emphasize the particular creep behavior of ODS alloys. To understand the influence of oxide particles, we focused on the dislocation microstructure and their interaction mechanisms with precipitates. Microstructural characterization were performed using Transmission Electron Microscopy (TEM) on both un-deformed alloys and on 650 °C–250 MPa creep tested MA957 samples. No noticeable differences were observed, and dislocations seem to be anchored by precipitate particles. The dynamic behavior was studied by in situ TEM straining experiments at room temperature on MA957. Observation of dislocation motions indicates that interactions with particles, including pinning, control the flow stress at least at room temperature. At 650 °C, other mechanism is probably predominant.

Simulation of self-point defects (SPDs) diffusion in elastic fields of edge and screw dislocations in slip systems 〈1 1 1〉{1 1 0}, 〈1 1 1〉{1 1 2}, 〈1 0 0〉{1 0 0}, 〈1 0 0〉{1 1 0} in the temperature range 293–1000 K has been performed by the object kinetic Monte-Carlo method for bcc iron and vanadium crystals with dislocation density ∼3 × 1014 m−2. Interaction energies of SPDs (elastic dipoles) with dislocations were calculated by means of the anisotropic theory of elasticity. These elastic interactions significantly change dislocation sink efficiencies. Dislocation sink efficiencies decrease as the temperature increases and tend to the limit value for linear sink without interaction field. The dislocations are more efficient sinks for self-interstitial atoms (SIAs) than for vacancies. The difference between the sink efficiencies for SIAs and vacancies is several times less for screw dislocations than for edge dislocations.

A study on zirconium nitride using TEM and nanoindentation was performed to assess the significant surface plasticity found to be introduced by sample polishing. Cross-sectional TEM results show strong evidence of plasticity via dislocations produced directly from surface grinding and polishing. These dislocations were found to glide on the { 0 1 1 } 〈 0 1 ¯ 1 〉 slip system. Using nanoindentation to observe the effects of surface dislocation density, a critical shear stress was found that relates to dislocations nucleation and multiplication. Continued chemo-mechanical polishing increased the critical shear stress to approximately 1600 mN by reducing surface dislocations. It is postulated that vacancy clusters and oxide microcrystallites produced during surface processing provide dislocation nucleation and/or multiplication sites. Gentle chemo-mechanical polishing for several hours greatly reduced or eliminated preexisting dislocations such that the critical shear stress for nucleation approaches the theoretical limit (∼G/5).

There are two creep regions with different creep characteristics: short-term creep region “H”, where precipitates and subgrains are thermally stable, and long-term creep region “L”, where thermal coarsening of precipitates and subgrains appear. In region “H”, the normalized subgrain size ( λ - λ 0 ) / ( λ ∗ - λ 0 ) has a linear relation with creep strain and its slope is 10 e −1 . But, region L is the time range in which the static recovery and the strain-induced recovery progress simultaneously. In this region, the static recovery accelerates the strain-induced recovery, and subgrain size is larger than that line which neglects the contribution of the static recovery. In region “L”, the Δ λ / Δ λ ∗ -strain present a linear relation with a slope 35 e −1 . There is a linear relation between hardness and subgrain size. Hardness drop, H 0 − H , as a function of Larson–Miller parameter can be a good measure method for assessment of hardness drop and consequently degradation of microstructure. Hardness drop shows an identical slope in creep region “H”, whereas hardness drop due to thermal aging and creep in region “L” show together a similar slope. In region “H”, degradation of microstructure is mainly due to recovery of subgrains controlled by creep plastic deformation, and precipitates do not have a major role. However, in creep region “L”, there are three degradation mechanisms that accelerate creep failure; (1) strain-induced recovery of subgrains due to creep plastic deformation, (2) static-recovery of subgrains and precipitates and (3) strain-induced coarsening of precipitates due to the appearance of static-recovery.

Ferritic steels strengthened by oxide particle dispersions (ODS) are prime candidate for future nuclear applications. So far, the beneficial mechanical characteristics of ODS steels are not fully understood, in terms of dislocation-based mechanisms. In this work, three-dimensional discrete dislocation dynamics simulations were carried out to analyze pre and post-irradiation plastic deformation in ferritic grains, with and without ODS particles. In the absence of irradiation induced defect loops, ODS-grains are stronger and plastic strain is more localized than in the corresponding, particle-free grain. After irradiation however, ODS-grain become more resistant to loop-induced hardening, while plastic strain spreading is broader, with respect to particle-free grain. This effect is due to dislocations accumulating next to the precipitates, generating internal stress allowing cross-slipped dislocations to go past the irradiation induced defect loops. Cross-slip is therefore a key feature of our model, for explaining the beneficial role of ODS particles on post-irradiation plastic deformation.

Normally, creep anisotropy of hcp metals is thought to be controlled by the crystallographic texture. Here, we show that the creep anisotropy of cold-worked Zr–2.5Nb tubes is also very dependent on the anisotropic dislocation structures introduced by cold-work. The contribution of each slip system to the creep deformation of an individual grain orientation depends upon the activity of that slip system during prior cold-work. This conclusion is reached by comparing the self-consistant visco-plastic polycrystalline models with thermal creep tests performed on internally pressurized thin-wall capsules with different textures under a transverse stress of 300 MPa at 350 °C, where dislocation creep is the dominant operating mechanism. The non-uniform dislocation distributions prior to creep were derived by simulating the cold-work process of Zr–2.5Nb tubes from an Elasto-Plastic Self-Consistent (EPSC) model.

14Cr–15Ni–Ti modified stainless steel alloyed with additions of phosphorus and silicon is a potential candidate material for the future cores of Prototype Fast Breeder Reactor. In order to optimise the titanium content in this steel, creep tests have been conducted on the heats with different titanium contents of 0.18, 0.23, 0.25 and 0.36 wt.% at 973 K at various stress levels. The stress exponents indicated that the rate controlling deformation mechanism was dislocation creep. A peak in the variation of rupture life with titanium content was observed around 0.23 wt.% titanium and the peak was more pronounced at lower stresses. The variation in creep strength with titanium content was correlated with transmission electron microscopic investigations. The peak in creep strength exhibited by the material with 0.23 wt.% titanium is attributed to the higher volume fraction of fine secondary titanium carbide (TiC) precipitates.

The impact of heavy-ion produced defects on the mobility of dislocations, dislocation sources and newly generated dislocations in 304 stainless steel are discovered by performing irradiation and deformation experiments in real time in the transmission electron microscope. Dislocations mobile prior to the irradiation are effectively locked in position by the irradiation, but the irradiation has no discernible impact on the ability of a source to generate dislocations. The motion and mobility of a dislocation is altered by the irradiation. It becomes irregular and jerky and the mobility increases slowly with time as the radiation-produced defects are annihilated locally. Channels created by dislocations ejected from grain boundary dislocation sources were found to have a natural width, as the emission sites within the boundary were spaced close together. Finally, the distribution of dislocations, basically, an inverse dislocation pile-up, within a cleared channel suggests a new mechanism for generating high local levels of stress at grain boundaries. The impact of these observations on the mechanical properties of irradiated materials is discussed briefly.

The correlation of defect energies with precipitation of the ferromagnetic phase near M23C6 carbide during creep tests at high temperature in Type 304 austenitic steel was examined by estimating the defect energies near the carbide, based on micromechanics. As one of the defect energies, the precipitation energy was calculated by assuming M23C6 carbide to be a spherical inclusion. The other defect energy, creep dislocation energy, was calculated based on dislocation density data obtained from transmission electron microscopy observations of the creep samples. The dislocation energy density was much higher than the precipitation energy density in the initial stage of the creep process, when the ferromagnetic phase started to increase. Creep dislocation energy could be the main driving force for precipitation of the ferromagnetic phase.

Anisotropic nucleation and growth of multi-classes of dislocation loops under the combined actions of fast-neutrons and an external applied stress are considered in modeling dislocation structure development in metals and alloys. The stochastic nature of the nucleation kinetics is formulated via the Fokker–Planck equation. The strain derived from the climb of the anisotropic dislocation structure is separable into volumetric and deviatoric components, corresponding respectively to swelling and creep. The creep contribution resulting from the development of the stress-induced dislocation anisotropy is found to be very significant and exhibits a strong correlation with swelling. For stainless steel, our model explains very well the complex deformation behavior observed in a wide variety of in-reactor experiments.

When a crystal deforms plastically, sources within, such as the Frank–Read source, emit dislocations, which then glide in response to the applied stress. As the dislocations move away from the source, they may encounter an obstacle, for example a grain boundary, impurity atom or locked dislocation, which they cannot overcome. As more dislocations are emitted, they ’pile up’ near the obstacle, until their own stress fields acting back on the source prevent more dislocations from being produced, unless the external applied stress is increased. The properties of these pile-ups strongly influence the deformation of the crystal as a whole giving rise to the Hall–Petch effect [E. Hall, Proc. Phys. Soc. B 64 (1951) 747–753, N. Petch, J. Iron and Steel Institute 173 (1953) 25–28.] relating yield strength to grain size. In this paper we investigate how the observed strong variation of elastic constants as functions of temperature affects the strength of interactions between dislocations, and mechanisms of plastic deformation of iron at elevated temperatures. We find that the observed severe softening of the tetragonal shear modulus C ′ at high temperature gives rise to a drastic reduction in the repulsion between parallel like edge dislocations, and hence to a greatly increased number of dislocations in pile-ups, especially those in the 〈1 0 0〉(0 0 1) configuration. The associated increase in plastic strain will lead in turn to a conspicuous reduction in tensile strength as C ′ falls. The phenomenon is inherently anisotropic, and this work underlines the importance of including anisotropic elastic effects when modelling Fe at high temperatures.

The in-reactor behaviour of internally pressurised capsules of Zr–2.5Nb tubes is analysed in detail to separate the stress dependent component of deformation (creep). It is found by a rigorous statistical analysis that the creep rate varies with fast neutron fluence. At 555 K the axial creep rate increases while the transverse creep rate decreases with fluence. At 588 K the creep rate in both the axial and transverse directions increases with fluence. It is also shown that the creep anisotropy ratio R , i.e., the ratio of axial to transverse creep rate for a pressurised tube, varies with fluence, stress and irradiation temperature. These findings are discussed in terms of the irradiation-induced evolution in microstructure. The possible impact of the evolution of the dislocation substructure is discussed with reference to a self-consistent polycrystalline model that takes into account the crystallographic texture and the grain interaction strains present in zirconium alloys. The lower temperature creep behaviour is consistent with an increase with fast fluence of the single crystal creep compliance related to prismatic dislocation climb and glide, or a decrease in the single crystal creep compliances relating to basal and pyramidal slip. The creep behaviour at the higher irradiation temperature is more complicated, and there may be an influence of phase changes as well as dislocation structure. It appears that all three eigenvalues describing the single crystal creep behaviour depend on fast fluence.

Pressurized thermal creep tubes of V–4Cr–4Ti have been examined following testing in the range 650–800 °C for tests lasting ∼10 4 h to provide comparison with tests on similar tubes following irradiation. It is found in all cases that creep results from dislocation motion. But the mechanism changes with increasing temperature and lower stress from one controlled by the climb and interaction of individual dislocations, to one controlled by sub-grain boundary structure that is created by relaxation of the interacting dislocations to lower energy planar arrays. This change in mechanism corresponds to a change from power law creep to Newtonian creep such that the stress exponent drops from ∼4 to ∼1. Although it is possible to explain the Newtonian response as Nabarro–Herring or Coble creep, it appears more likely that behavior is due to Harper–Dorn creep, in which case the change in response occurs at the Peierls stress.

Pressurized thermal creep tubes of highly purified V–4Cr–4Ti, the NIFS-Heat-2 alloy have been examined following testing in the range 700–850 °C. It was found that the creep stress exponent of the NIFS-Heat-2 alloy is about five and that the characteristic creep mechanism was the dislocation creep usually observed in pure metals. The apparent activation energy of creep deformation is about 210 kJ/mol in the temperature range 700–850 °C. Creep deformation was considered to be controlled by climb-controlled dislocation-glide at 850 °C, where sub-grain boundary structure predominates and consists of dislocation dipole structures and pile-ups of dislocations.

To understand the channeling deformation mechanism in irradiation-damaged materials, the interaction processes between dislocations and radiation induced defect clusters was investigated by molecular dynamics simulation. Small interstitial Frank loops can be easily eliminated by dislocation, but larger ones can stay at original sites after intersection. Vacancy Frank loops are weaker obstacle to dislocation compared with the interstitial ones and they cannot be absorbed by dislocation at low temperature.

We calculated the formation energy and the migration energy of vacancies near perfect edge dislocations in Cu. Effective vacancy volume in the most compressive site (the dislocation core) was 94.3% of that in the perfect lattice. The vacancy formation energy only in the most compressive site was conspicuously low (0.74 eV), and the energy in other sites was almost equal to that in the perfect lattice. As the vacancy got close to the dislocation core from the compression side, the vacancy migration energy in the direction of the Burgers vector and also on the extra half plane decreased. Calculated migration energies suggest the trapping of vacancies on the compression side of the edge dislocation and the de-trapping at high temperatures.

SUS 316 stainless steel will be used in ITER as a major structural material. However, accumulation of radiation-induced deformation data in the anticipated operation temperature range, centered at about 300 °C, is not adequate. Although it will be used mainly in annealed condition, cold-working and consequent introduction of network dislocations is inevitable at various parts of the structure. Computer simulation was performed to examine the radiation-induced deformation of SUS 316 steel, with emphasis on the effect of network dislocation density, for the ITER-relevant condition. Irradiation creep rates and radiation-induced stress relaxation behaviors were examined at 300 °C for various displacement rates, applied stresses and network dislocation densities. Rather complicated dependence on dislocation density has been predicted in both creep deformation and stress relaxation. This complexity results from a competitive redistribution of point defects to the operating deformation mechanisms and from the predominance among them that depends on the network dislocation density.

In the climb-enabled glide model of irradiation creep, the plastic deformation is defined by the elastic deflections of pinned dislocations, which is an inconsistency. We argue that this relation is incorrect; instead, as in other pinning-unpinning-type models, the dislocations move from one set of obstacles to another, so that the inter-obstacle spacing determines creep rate, whereas the dependence on the applied stress is only implicit in the unpinning time.

High temperature performance tests including tensile, creep and creep-fatigue tests were conducted using specimens from a large ingot of highly purified V–4Cr–4Ti alloys, NIFS-Heat1. The obtained tensile data of NIFS-Heat1 alloys agree well with previous V–4Cr–4Ti alloys data. The most significant effect for the impurity reduction in NIFS-Heat1 alloys was the higher temperature for the appearance of the dynamic strain hardening compared to previous heats of V–4Cr–4Ti. From uniaxial creep tests, a creep-stress exponent of the NIFS-Heat1 alloy is about 5–8 and the characteristic of creep mechanism is the dislocation creep in pure metals. The apparent activation energies of creep deformation ranged from 180 to 210 kJ/mol at a temperature regime from 750 to 800 °C. No significant degradation due to fatigue behavior was observed during creep-fatigue tests. The creep-fatigue behavior was superior to the load-constant creep behavior in the same stress and temperature conditions at a lower stress condition.

The physical framework employed for most of the current irradiation creep models of cubic polycrystalline materials is a single crystal with a dislocation system consisting of three orthogonal Burgers vectors. The irradiation creep models based on this crystal system can be broken into two categories, one where irradiation creep occurs by dislocation climb alone, and the other where irradiation creep occurs by a combination of dislocation climb and glide. Often, irradiation creep is presumed to proceed by dislocation climb alone at low applied stresses, but by a combination of dislocation climb and glide at higher applied stresses. However, the rules for generalized deformation, as applied to dislocation climb, determine if a polycrystal is capable of deformation by dislocation climb alone. The current work analyzes generalized deformation by dislocation climb of fcc and bcc stainless steel crystals. Some implications for irradiation creep of polycrystalline fcc and bcc stainless steels are discussed.

The onset of plastic instability in neutron irradiated copper is investigated by computer simulation of the dynamics of the elastic interaction between dislocation loops emitted from Frank–Read (F–R) sources and irradiation-induced defect clusters. We show that small prismatic defect clusters produced directly from collision cascades are trapped in the stress field of slip dislocations, and that mobile clusters are absorbed in the dislocation core, when they approach within ∼6 nm. Sessile vacancy clusters are also absorbed within this ‘stand-off’ distance because of an induced surface tension on their stacking fault. The interaction between prismatic defect clusters in ‘decorations’ and dislocations is shown to provide significant resistance to the initiation of plastic deformation in irradiated copper (source hardening). Sessile stacking fault tetrahedra are also shown to resist dislocation motion by localized forces before they are absorbed and removed by activated dislocation sources. The significance of these mechanisms to initiation of localized deformation and plastic instability are discussed.

A dislocation dynamics (DD) model for plastic deformation, connecting the macroscopic mechanical properties to basic physical laws governing dislocation mobility and related interaction mechanisms, has been under development. In this model there is a set of critical reactions that determine the overall results of the simulations, such as the stress-strain curve. These reactions are, annihilation, formation of jogs, junctions, and dipoles, and cross-slip. In this paper we discuss these reactions and the manner in which they influence the simulated stress- strain behavior in fcc and bcc metals. In particular, we examine the formation (zipping) and strength of dipoles and junctions, and effect of jogs, using the dislocation dynamics model. We show that the strengths (unzipping) of these reactions for various configurations can be determined by direct evaluation of the elastic interactions. Next, we investigate the phenomenon of hardening in metals subjected to cascade damage dislocations. The microstructure investigated consists of small dislocation loops decorating the mobile dislocations. Preliminary results reveal that these loops act as hardening agents, trapping the dislocations and resulting in increased hardening.

In this work, based on results of the tests with study of dynamic diagrams of compression of uranium–molybdenum alloy, we made an attempt to determine dislocation velocity, length of free run of dislocations, and increase of dislocation density during plastic deformation of alloy at dynamic strain rates.

Dynamic behavior of small interstitial clusters (dislocation loops), such as the interaction with a self-interstitial atom (SIA) in Fe and the interaction between a SIA and an edge dislocation in Fe have been studied by means of molecular dynamics (MD) method in order to clarify their role in the evolution of damage structure during irradiation, especially in the so-called production bias effect through one-dimensional migration to sinks and in the so-called dislocation bias. Model crystal was constructed by using N-body potentials and small interstitial clusters, i.e., bundle of crowdions, a SIA and an edge dislocation were inserted. It was found that a SIA migrates to the edge dislocation core and finally the direction of the crowdion is converted to the direction parallel to the Burgers vector of the edge dislocation. Dynamic behavior of the loop, e.g., the interaction with a crowdion on a central 〈1 1 1〉 loop axis was also investigated as a function of time. It was shown that small initial cluster, e.g., I 19 in Fe is very mobile under the interaction with a crowdion, which shows that this interstitial cluster I 19 has already property of a dislocation loop of edge character and low Peierls potential for the motion of this loop, which is consistent with the straight edge dislocation in Fe.

Zircaloy-4 cladding tubes with different area reductions at the final cold working were subjected to creep testing at temperatures of 350 to 500°C. The creep testing was carried out mainly at the biaxial stress state. The creep rate and creep strain of the Zircaloy-4 cladding tube increased with increasing area reduction at the final cold working. The creep model was derived to cover the effect of area reduction and then verified by using supplementary creep data obtained at 350°C and 138 MPa. Furthermore, this generalized creep model was verified to well describe the thermal creep of Duplex tubes and low Sn Zircaloy-4 tubes. The creep rate of Zircaloy-4 cladding tubes followed the exponential stress dependence, not the power law creep. The creep activation energy was determined to be 60 kcal/mol, leading to the conclusion that creep is controlled by dislocation creep. Thus, it is suggested that the increase of creep rates and creep strains with the increase in cold working is due to an enhanced climb rate arising from the increased concentration of vacancy.

In order to obtain information on the bias factor in void swelling in metals, computer simulation in a model lattice has been performed for the interaction between a dislocation, or a dislocation loop and point defects in both fcc (Au) and bcc (Fe) metals. Not only a straight dislocation but also a jogged dislocation was chosen for the calculation of the interaction. In fcc metal, the calculation was made for the interaction between an extended dislocation and point defects. The effective interaction range, L, was determined for all the cases, and from these values the so-called bias factor was obtained as 55 and 19% for fcc (Au) and bcc (Fe) metals, respectively in the case of a straight dislocation. Almost no difference for this range L was seen between a dislocation without a jog and that with a jog. However, sink efficiency is considered to be higher for a jogged dislocation than a straight one. A comparatively small bias factor, about 8%, was obtained for the case of a dislocation loop and point defects in the fcc lattice.

According to our model on the mechanism of dislocation bias reduction based on the interaction of dumbbell self-interstitial atoms (SIAs) with dislocation, the bias is significantly different depending on the dumbbell configuration in the dislocation strain field. A large-scale molecular dynamics (MD) simulation is performed to reveal the stability and the mechanism of diffusion of dumbbell SIAs near the edge dislocation core in bcc iron. Most SIAs take the crowdion configuration parallel to the Burgers vector in the expansion side of the dislocation. Such crowdions are stable in the temperature range of this simulation, i.e. between 373 and 473 K, making one-dimensional random to-and-fro motion parallel to the dislocation Burgers vector staying at several atomic layers “below” the dislocation core. This means that the SIA does not approach the dislocation core. These results suggest that the stable configuration of SIAs is seriously affected by the dislocation resulting in a reduction of bias factor.

Point defect currents to straight edge dislocations in fcc and bcc lattices are calculated taking into account the strain-induced diffusion anisotropy (elastodiffusion) of point defects in combined strain fields from dislocations and external loads. The dislocation bias factors are shown to depend on dislocation orientation with respect to the lattice symmetry axes and the direction of external load. The implications of the orientational dependence of bias factors for the dislocation loop kinetics, swelling and irradiation creep in cubic lattices are discussed.

The indentation method was found to be a suitable method for determination of the creep properties of small weld joints of nuclear fuel pins. The activation energy for the creep for TIG welded sample was found to be ~40% more than the base metal. Both base metal and resistance-welded specimens have the same activation energy for creep and match with self-diffusion values of Zricaloy. Since the value of n estimated in the study is in the range of 5.5-7.6 the most probable mechanism is dislocation creep in the power-law regime.

Results of a simple creep test of three FE-12Cr-20Mn-1W alloys are presented, performed at temperatures from 873 to 973 K and at applied stresses corresponding to dislocation creep conditions. The influence of the carbon content on creep behaviour is established. The creep behavior of the steels is also comparedwith that of some standdard Fe-Cr-ni and Fe-Cr-Mn-Ni austenitic stainless steels.

The effect of external loading on vacancy pipe diffusion along a dislocation line is studied. It is shown that the external load leads to orientational dependence of dislocation sink strengths due to polarization of vacancies in the dislocation core. Computer calculations of the dipole tensor for vacancy diffusion in the core of a [100](010) dislocation in α-Fe are performed to obtain an estimate of the irradiation creep compliance value.

A model of plastic deformation evolution and flow localisation processes in irradiated materials is proposed. This models takes into account the dislocation distribution function dependence on dislocation velocity in an ensemble. It is shown that the fraction of dislocations overcoming radiation defects with high velocities in the dynamical regime grows with increasing radiation hardening.

Annealed type-348 stainless steel specimens irradiated to 33 to 39 dpa at 350°C were examined by transmission electron microscopy to determine the cause of pronounced irradiation creep and hardening. The irradiation produced very high densities of 1–2 nm diameter helium bubbles, 2–20 nm diameter faulted (Frank) dislocation loops and 10 nm diameter precipitate particles. These defects account for the observed irradiation hardening but do not explain the creep strains. Too few point defects survive as faulted dislocation loops for significant creep by the stress-induced preferential absorption (SIPA) mechanism and there are not enough unfaulted dislocations for creep by climb-induced glide. Also, the irradiation-induced precipitates are face-centred cubic G-phase (a niobium nickel suicide), and cannot cause creep. It is suggested that the irradiation creep occurs by a grain-boundary movement mechanism such as diffusion accomodated grain-boundary sliding.

Creep by climb-enabled glide of dislocations may be calculated by mathematically specifying both the origin of the net point defect flux producing the climb and the mode of glide deformation. In general, the net flux contains components arising from (1) differences in point defect capture efficiencies among dislocations and other types of sinks such as cavities, and (2) from differences in capture efficiencies among dislocations whose Burgers vectors are at different orientations with respect to the stress direction. Proposed models based on the first component and employing several modes of glide deformation are reviewed. The characteristics of creep based on the second component, stress induced preferred absorption of point defects on dislocations and subsequent glide, PAG-creep, are described for one mode of glide deformation. Effects of point defect trapping at impurities on creep by this mechanism and on the corresponding climb component of creep are evaluated. Comparison is made with experimental data from light-ion irradiations. The sensitivity of predicted creep rates to the detailed microstructure is emphasized.

Vacancy and interstitial interactions with the straight edge dislocation 1 2 [111] (1 1 0) in α-Fe have been investigated by a computer simulation method. Zones of spontaneous absorption of point defects by the edge dislocation have been determined. The mean zone radii values were 1.185 nm and 0.340 nm for an interstitial and a vacancy, respectively. The main mechanism of spontaneous absorption has been revealed: the initiation of the replacement chain by the interstitial followed by attachment of the chain's last atom to the dislocation extra half-plane edge. Bond energies of the dislocation with point defects in the absorption zones were about 4 eV and 1 eV for the interstitial and the vacancy, respectively, and significantly higher for those beyond these zones. The effect of interstitial configuration and orientation on the interaction with the dislocation has been evaluated. Application of uniaxial tension to a stress comparable to the flow stress did not affect the sizes of absorption zones. The sizes changed at higher stress levels (5 or 25 times). The relevence of the results to radiation-damage studies is discussed.