Final Report on Neutron Irradiation at Low Temperature to Investigate Plastic Instability and at High Temperature to Study Cavitation (original) (raw)
Related papers
Nuclear Materials and Energy, 2020
In this work, we used Dislocation Dynamics (DD) simulations to investigate the role of the hierarchical defects microstructure of ferritic-martensitic steel Eurofer97 in determining its hardening behavior. A Representative Volume Element (RVE) for DD simulation is identified based on the typical martensitic lath size. Material properties for DD simulations in b.c.c Eurofer97 are determined, including the dislocation mobility parameters. The dependence of material parameters on temperature is fitted to experimental yield strength measurements carried out at room temperature and 330 • C, respectively. Voids and precipitates observed in the microstructure, such as M 23 C 6 and Tantalum-rich MX, are considered in our DD simulations as inclusions with realistic size distribution and volume density, while 〈1 1 1〉-and 〈1 0 0〉-type irradiation loops are included directly in the DD simulations. The lath structure, together with its typical precipitates arrangement and the different crystallographic orientation of the martensitic blocks can also be captured in the simulations. DD simulations are used to extract microstructure-specific hardening parameters, which can be used to simulate the properties of Eurofer97 at the engineering scale.
Journal of Nuclear Materials, 2005
The REVE project (REactor for Virtual Experiments) is an international effort aimed at developing tools to simulate irradiation effects in light water reactors materials. In the framework of this project, a European team developed a first tool, called RPV-1 designed for reactor pressure vessel steels. This article is the third of a series dedicated to the presentation of the codes and models used to build RPV-1. It describes the simplified approach adopted to simulate the irradiation-induced hardening. This approach relies on a characterization of the interactions between a screw dislocation and irradiation-induced defects from molecular dynamics simulations. The pinning forces exerted by the defects on the dislocation were estimated from the obtained results and some hypotheses. In RPV-1, these forces are used as input parameters of a Foreman and Makin-type code, called DUPAIR, to simulate the irradiation-induced hardening at 20°C. The relevance of the proposed approach was validated by the comparison with experimental results. However, this work has to be considered as an initial step to facilitate the development of a first tool to simulate irradiation effects. It can be improved by many ways (e.g. by use of dislocation dynamics code).
Dislocation decoration and raft formation in irradiated materials
Philosophical Magazine, 2005
Experimental observations of dislocation decoration with self-interstitial atom (SIA) clusters and of SIA cluster rafts are analysed to establish the mechanisms controlling these phenomena in bcc metals. The elastic interaction between SIA clusters, and between clusters and dislocations is included in kinetic Monte Carlo (KMC) simulations of damage evolution in irradiated bcc metals. The results indicate that SIA clusters, which normally migrate by 1D glide, rotate due to their elastic interactions, and that this rotation is necessary to explain experimentally-observed dislocation decoration and raft formation in neutronirradiated pure iron. The critical dose for raft formation in iron is shown to depend on the intrinsic glide/rotation characteristics of SIA clusters. The model is compared with experimental observations for the evolution of defect cluster densities (sessile SIA clusters and nano-voids), dislocation decoration characteristics and the conditions for raft formation.
Microstructural analysis of deformation in neutron-irradiated fcc materials
Journal of Nuclear Materials, 2006
Plastically deformed microstructures in neutron-irradiated face centered cubic (fcc) materials, copper, nickel, and 316 stainless steel (316SS), were investigated by transmission electron microscopy (TEM). Neutron irradiation in the range of 65-100°C up to 1 displacement per atom (dpa) induced a high number density of black spots, stacking fault tetrahedra (SFT) and Frank loops, which resulted in irradiation-induced hardening. Deformation of irradiated fcc materials induced various microstructures, such as dislocation channels, stacking faults, and twins. In the 316SS irradiated to 0.1-0.8 dpa, the deformation microstructure consisted of a mixture of dislocation bands, tangles, twins, dislocation channels, and also martensite phase. Deformation-induced martensite transformation tends to occur with dislocation channeling, suggesting that localized deformation could lead to transformation of austenite to martensite at a high stress level. At higher irradiation doses (0.1-1 dpa), dislocation channeling became the dominant deformation mode in fcc materials, and is coincident with prompt plastic instability at yield. The channel width seems to be wider when the angle between tensile direction and dislocation slip direction is close to 45°. Furthermore, the correlation between channel width and resolved shear stress appears to be material dependent, with copper having the greatest slope and 316SS the smallest.
Modeling dislocation evolution in irradiated alloys
Metallurgical Transactions A, 1990
Neutron irradiation of structural materials leads to such observable changes as creep and void swelling. These effects are due to differential partitioning of point defects. Although most radiationproduced point defects recombine with an antidefect, a very small fraction of the defects survives. The surviving defect fraction is directly related to the density and type of extended defects that act as point defect sinks. Defect partitioning requires the presence of more than one type of sink and that at least one of the sinks has a capture efficiency for either vacancies or interstitials that is different from that of the other sink(s). For example, dislocations provide the interstitial "bias" that drives swelling, and the ratio of the dislocation to cavity sink strength determines the swelling rate. These sink strengths change during irradiation, and an explicit model of their evolution is required to simulate swelling or creep. Such a model has been developed; the influences of various model assumptions and parameters are discussed. The model simulates the evolution of Frank faulted interstitial loops, providing a dislocation source and the glide/climb of the dislocation network leading to annihilation of dislocation segments. Good agreement is found between model predictions and experimental data. Swelling simulations are shown to be quite sensitive to the dislocation model.
Metals, 2020
Several open issues remain concerning the quantitative understanding of irradiation hardening in high-Cr steels. One of these issues is addressed here by correlating yield points that are observed in stress-strain curves with dislocation decoration observed by TEM for neutron-irradiated Fe-Cr alloys. It is found that both higher neutron exposure and higher Cr content promote irradiation-induced loops to arrange preferentially along dislocation lines. Consequently, the activation of dislocation sources requires unlocking from the decorating loops, thus resulting in a yield drop. This process is considered within the source hardening model as opposed to the dispersed barrier hardening model, the latter aimed to describe dislocation slip through a random array of obstacles. Microstructure-informed estimates of the unlocking stress are compared with measured values of the upper yield stress. As functions of neutron exposure, a cross-over from the dominance of dispersed-barrier hardening...
Acta Materialia, 2012
Low temperature irradiation of crystalline materials is known to result in hardening and loss of ductility, which limits the usefulness of candidate materials in harsh nuclear environments. In body-centered cubic (bcc) metals, this mechanical property degradation is caused by the interaction of in-grown dislocations with irradiation defects, particularly small dislocation loops resulting from the microstructural evolution of displacement cascades. In this paper, we perform dislocation dynamics simulations of bcc Fe containing various concentrations of dislocation loops produced by irradiation in an attempt to gain insight into the processes that lead to hardening and embrittlement. We find that a transition from homogenous to highly localized deformation occurs at a critical loop density. Above it, plastic flow proceeds heterogeneously, creating defect-free channels in its wake. We find that channel initiation and size are mediated by loop coalescence resulting from elastic interactions with moving dislocations.
Molecular dynamics modeling of cavity strengthening in irradiated iron
Journal of Computer-Aided Materials Design, 2007
One of the most important problems in the field of nuclear industry is the relationship between irradiation-induced damage and the resulting induced mechanical response of the target metal and in particular ferritic base steels. In this work molecular dynamics simulation is used to simulate the nanoscale interaction between a moving dislocation and a defect, such as a cavity, as void or He bubble. The stress-strain curves are obtained under imposed strain rate condition using the atomic potentials based on the Fe potential of Ackland et al. 1997 for a void and He bubble as a function of He content and temperature. It appears that a 2 nm void is a stronger obstacle than a He bubble at low He contents, whereas at high He contents, the He bubble becomes a stronger obstacle. With increasing temperature the escape stress decreases and at the same time there is increasing degeneracy in the type of interaction.
Microstructure evolution of T91 steel after heavy ion irradiation at 550 °C*
Chinese Physics B, 2021
Fe-Cr ferritic/martensitic (F/M) steels have been proposed as one of the candidate materials for the Generation IV nuclear technologies. In this study, a widely-used ferritic/martensitic steel, T91 steel, was irradiated by 196-MeV Kr+ ions at 550 °C. To reveal the irradiation mechanism, the microstructure evolution of irradiated T91 steel was studied in details by transmission electron microscope (TEM). With increasing dose, the defects gradually changed from black dots to dislocation loops, and further to form dislocation walls near grain boundaries due to the production of a large number of dislocations. When many dislocation loops of primary a 0/2〈 111 〉 type with high migration interacted with other defects or carbon atoms, it led to the production of dislocation segments and other dislocation loops of a 0 〈 100 〉 type. Lots of defects accumulated near grain boundaries in the irradiated area, especially in the high-dose area. The grain boundaries of martensite laths acted as imp...
Journal of Nuclear Materials, 2008
Hardening and embrittlement are controlled by interactions between dislocations and irradiation induced defect clusters. In this work we employ the visco plastic self consistent (VPSC) polycrystalline code in order to model the yield stress dependence in ferritic steels on the irradiation dose. We implement the dispersed barrier hardening model in the VPSC code by introducing a hardening law, function of the strain, to describe the threshold resolved shear stress required to activate dislocations. The size and number density of the defect clusters varies with the irradiation dose in the model. We find that VPSC calculations show excellent agreement with the experimental data set. Such modeling efforts can both reproduce experimental data and also guide future experiments of irradiation hardening.