Irradiation creep of engineering alloys in nuclear reactor cores differs from the creep that is observed outside of the irradiation environment. It exhibits characteristics like high temperature thermal creep because it occurs in an environment of elevated vacancy point defect concentrations, but one must also consider the effect of interstitial point defects and the effect of both vacancy and interstitial concentrations, which are greater than the thermal equilibrium values, on an evolving microstructure. Irradiation creep is dependent on the point defect flux to different sinks and can be modelled using conventional rate theory. The net interstitial or vacancy point defect flux to different sinks determines the strain rate in a direction that can be considered perpendicular to the plane of the sink, which is the extra half plane of an edge dislocation or the plane of a grain boundary. There has been increasing evidence that, for complex alloys such as Zr-2.5Nb pressure tubing in CANDU reactors, the irradiation creep is largely dependent on the grain structure (size and shape). While the maximum amount of thermal creep by dislocation slip will be proportional to the distance a dislocation travels, i.e., proportional to the grain dimension in the direction of slip, observations indicate that the magnitude of irradiation creep is inversely proportional to the grain dimensions, indicating a creep mechanism dependent on diffusional mass transport. Mechanistic modelling of irradiation creep based on rate theory is described and used to account for high diametral creep rates observed for pressure tubes with unusual microstructures fabricated by non-standard fabrication routes.

A competing-rates model is presented to account for operational changes in the metastable β-Zr phase of the Zr-2.5Nb alloy used to make CANDU reactor pressure tubes and is used to predict temperature gradients at the outlet rolled joints using the decomposition of the β-Zr phase as a proxy for temperature. High temperatures decompose the β phase by enhancing the formation of small particles of ω and α phases. Fast neutron flux causes the ω and α phases to shrink. This process is assumed to depend on the total volume of the particles, because they are comparable to, or smaller than, the size of the neutron displacement cascades. The barrier energy for thermal growth was determined to be 2.43 eV, when an Arrhenius A factor of 1013/s was assumed. The cross section for (ω+α)-phase shrinkage is 24.5 barns for Zr-2.5Nb irradiated in CANDU reactors. Assuming that the shrinkage is dominated by the migration of self-interstitial point defects, a defect production efficiency of 1.4% was found.

The evolution of the mechanical properties of Zr-2.5Nb pressure tubing during irradiation is dependent on dislocation loop densities that are represented by the broadening of X-ray diffraction lines. Empirical models for the integral breadth of the diffraction peaks as a function of operating conditions have been developed to predict the mechanical properties of CANDU reactor pressure tubes as a function of fast neutron flux, time and temperature. Apart from predicting mechanical property changes based on integral breadth measurements, a new model has been developed to retrospectively deduce abnormal operating temperatures of ex-service pressure from the measured line broadening. The application of integral breadth measurements to assess mechanical properties and temperature variations in pressure tubes is described and discussed in terms of the implications for pressure tube integrity.

The diffraction line profiles of single crystal Zr and Zr-2.5Nb pressure tubing have been measured and analysed before and after irradiation in a nuclear reactor. The irradiation produces damage in the form of prismatic dislocation loops. In all cases, Fourier analysis of the deconvoluted peak using the method of Warren and Averbach shows that the magnitude of the distortion (micro-strain) component is not consistent with the dislocation density measured by transmission electron microscopy. The discrepancy can be related to the reduced strain energy from the dipoles on opposite sides of the prismatic dislocation loops and loop alignment, which reduces the average strain in the same way as polygonization or sub-grain boundary formation. The Fourier analysis also gives a value for the coherent diffraction domain size that is consistent with the mean spacing between the dislocations and thus the dislocation density.

The processes that control irradiation creep are dependent on the temperature and the rate of production of freely migrating point defects, affecting both the microstructure and the mechanisms of mass transport. Because of the experimental difficulties in studying irradiation creep, many different hypothetical models have been developed that either favour a dislocation slip or a mass transport mechanism. Irradiation creep mechanisms and models that are dependent on the microstructure, which are either fully or partially mechanistic in nature, are described and discussed in terms of their ability to account for the in-reactor creep behaviour of various nuclear reactor core materials. A rate theory model for creep of Zr-2.5Nb pressure tubing in CANDU reactors incorporating the as-fabricated microstructure has been developed that gives good agreement with measurements for tubes manufactured by different fabrication routes having very different microstructures. One can therefore conclude that for Zr-alloys at temperatures < 300 °C and stresses < 150 MPa, diffusional mass transport is the dominant creep mechanism. The most important microstructural parameter controlling irradiation creep for these conditions is the grain structure. Austenitic alloys follow similar microstructural dependencies as Zr-alloys, but up to higher temperature and stress ranges. The exception is that dislocation slip is dominant in austenitic alloys at temperatures < 100 °C because there are few barriers to dislocation slip at these low temperatures, which is linked to the enhanced recombination of irradiation-induced point defects.

Inconel & REG; X-750 has been implanted with helium at a temperature of 400 degrees C to obtain an average concentration of [He] of about 3000 atomic parts per million (appm). The implantation resulted in the formation of a low density of point defect clusters in the form of dislocation loops and a high density of He-stabilized cavities. Uni-axial tensile tests were conducted at room temperature on both the helium-implanted and non-helium-implanted materials. There was a reduction in ductility and an increase in yield strength after helium implantation. The fracture surface of the helium-implanted-material exhibited a mixed failure mode of inter-granular and transgranular fracture. TEM observations showed that the trans-granular fracture is the result of failure along twin/martensite platelets that were present in the material prior to implantation. The interfaces of these platelets with the austenite matrix, which are bordered either by incoherent twinned material or thin layers of & epsilon;-martensite, are paths for crack propagation that may be enhanced by He segregation at the interfaces. The results are discussed in terms of the implications for observations of channel fracture in neutron irradiated austenitic stainless steels and Ni-alloys.

The production of prismatic dislocation loops in nuclear reactor core materials results in hardening because the loops impede dislocation motion. Yielding often occurs by a localised clearing of the loops through interactions with gliding dislocations called channeling. The cleared channels represent a softer material within which most of the subsequent deformation is localized. Channeling is often associated with hypothetical dislocation pileup and intergranular cracking in reactor components although the channels themselves do not amplify stress as one would expect from a pileup. The channels are often similar in appearance to twins leading to the possibility that twins are sometimes mistakenly identified as channels. Neither twins nor dislocation channels, which are bulk shears, produce the same stress conditions as a pileup on a single plane. At high doses, when cavities are produced (either He-stabilised bubbles at low temperatures or voids at high temperatures), there can be reduced ductility because the material is already in an equivalent advanced stage of microscopic necking. He-stabilised cavities form preferentially on grain boundaries and at precipitate or incoherent twin/ε-martensite interfaces. The higher planar density of the cavities, coupled with the incompatibility at the interface, results in a preferential failure known as He embrittlement. Strain localisation and inter- or intragranular failure are dependent on many factors that are ultimately microstructural in nature. The mechanisms are described and discussed in relation to reactor core materials.

An accelerated irradiation method using ion irradiation of recrystallized and precipitation-hardened Inconel X750 has been used to study the effect on mechanical properties and microstructure of: (i) atomic displacement damage; and (ii) helium implantation. The atomic displacement damage was produced by irradiation with 5 MeV protons at 400 degrees C and varied from about 0.1 dpa at the surface to 1 dpa at a depth of about 75 mu m in a tensile test specimen that was 80 mu m thick. Helium implantation was conducted at 450 degrees C using a range of He ion energies to obtain a relatively uniform distribution within an 80 mu m thick tensile specimen with peak levels of about 3000 appm helium. The tensile specimens were tested in uniaxial tension at room temperature using a conventional load frame to separately determine the effect of irradiation damage and helium-implantation on mechanical properties. For both cases the yield and ultimate tensile strength increased, and the ductility decreased, even though the atomic displacement damage and [He] concentrations were low and non-uniform. TEM examination showed that the increased strength is attributed to point defect clusters in the form of loops for the proton irradiated material and He-stabilized cavities for the He-implanted material.

It is a widely held belief that dislocation slip has a direct effect on crystal orientation. Some of the confusion may be attributed to semantics when researchers are referring to related effects of dislocations on crystal orientation; either elastic bending due to constraints or the creation of geometrically necessary dislocations by climb. This communication highlights the distinction between the two and discusses why what is often imagined conflicts with what is real and possible. It is demonstrated that deformation-induced changes in the orientation of crystals are primarily limited to twinning and collections of geometrically necessary dislocations (GNDs), which in the most extreme cases are sub-grain boundaries. Alternate explanations for texture changes related to dislocation slip are provided, and they challenge the notion that grains can simply rotate because of dislocation slip through some undefined mechanism.

Zirconium alloys have widespread applications in nuclear energy, with Zr-2.5Nb commonly being used as pressure tube material in reactors. Their microstructure encompasses intermetallic nanoprecipitates (NPs) and solutes that significantly impact their behavior in corrosive environments and irradiation. Hence, we analyze the crystal structure of Zr((Zr,) Nb,Fe)2 NPs using transmission electron microscopy (TEM), electronic density functional theory (DFT) calculations, and finite element analysis (FEA). Our findings unveil a mixed c14 and c15 Laves phase structure within the NPs and provide an explanation through the syncroshear mechanism. Through thermodynamic analysis, we evaluate the electronic, vibrational, and strain contributions to the free energy of the NPs. Our results indicate that the c15 structure is energetically favored at temperatures below 600 K, while the c14 structure prevails at higher temperatures. We provide an explanation for the observed coexistence of these structures in the NPs based on two key insights: (1) During annealing at high temperatures, the energetically favorable c14 NPs form, and (2) as the alloy cools, a partial phase transition to the c15 structure occurs, constrained by kinetic limitations. Furthermore, our study reveals that the NP/a-Zr interface is likely to be incoherent due to the considerable stresses involved. This finding is consistent with high-resolution transmission electron microscopy (HRTEM) micrographs, which demonstrate the presence of an incoherent interface.

Rate theory models have been developed for the swelling and He-embrittlement of austenitic stainless steels and Ni-alloys in nuclear reactors. The models illustrate how microstructure evolution during irradiation affects the rate of change of mechanical properties and the dimensional stability. He-stabilised cavity accumulation on grain boundaries, which causes brittle failure at low stresses and strains known as He-embrittlement, is shown to be strongly dependent on the irradiation temperature and the rate of production of Frenkel pairs and He atoms. The results show that the accumulation of cavities on grain boundaries falls into two regimes: (i) that dictated by matrix bubble swelling at low temperatures; and (ii) that dictated by matrix void swelling at high temperatures.

Aging of Inconel X-750 spacers in Canada Deuterium Uranium (CANDU) reactors is managed through a combined program of in-service inspection, component removal for material surveillance testing, research and development, and fitness-for-service evaluation. Crush tests of exservice spacer coils have shown that the load carrying capacities of Inconel X-750 spacers decrease with operating time. The spacers also exhibit reduced ductility and inter-granular failure that is typical of helium (He)-embrittlement. Examination of ex-service spacers confirmed that material interfaces (mainly grain boundaries) were perforated by He-bubbles. Assuming that the grain boundary bubble coverage dictates the grain boundary strength and thus the failure load of the polycrystalline component, we have simulated the degradation and inter-granular failure of the spacer coils at low and high operating temperatures. The observed inter-granular fracture has been simulated by a cohesive zone model that governs the grain boundary de-bonding process. The grain boundary strength is related to the perforation by He-stabilised cavities. The model of grain boundary traction separation consists of three parts: (i) a rate-theory calculation to predict grain boundary coverage; (ii) a local material degradation model to quantify helium induced grain boundary strength reduction; (iii) a traction separation model to simulate grain boundary de-cohesion. The results show that the accumulation of Hestabilised cavities on grain boundaries causes material failure at lower stresses and strains that are dependent on the irradiation temperature and the neutron exposure. The load carrying capacity of the X-750 spacers decreases non-linearly with irradiation and the rate of change is decreasing with increasing dose.