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2/28/2007 5:24 PM
Multiscale Modeling of Helium Transport and Fate in Irradiated Nanostructured Ferritic Alloys
The purpose of this work is to develop and verify multiscale models of He transport for candidate fusion alloys. Alloy design is focused on breaking up He clusters into fine-scale bubbles. Specifically, a current focus is to integrate MD and kMC models into a rate-theory code in order to reach a multi-scale level for truly predictive capability. Results show that at 500°C in tempered martensitic steel (TMS) most bubbles form in dislocation junctions and also along the grain boundary; the matrix is not a favored site for bubble nucleation and growth. However, with the presence of precipitates, while bubble nucleation still takes place initially at dislocations, most nucleation and growth occurs at precipitates interfaces. In other words, nano precipitates act as efficient He clustering sites. Increasing the dislocation density results in a higher number density of He bubbles with a corresponding decrease in the average radius of the bubbles. Increasing the He/dpa ratio has a similar effect, increasing density and decreasing size.
Helium-vacancy Cluster Evolution during Hot Helium Implantation in Nickel
S. I. Golubov
There are many mechanisms that must be considered when trying to determine the clustering evolution of He under radiation. This includes interstitial and vacancy mechanisms for diffusion as well as substitutional diffusion. While there are many ways to simulate homogeneous clustering, only Object kMC can treat the problem non-homogeneously.
In considering He clustering using rate theory, vacancy and He clustering must be treated separately in order to accurately determine the clustering. In particular, small nano-sized vacancy and He clusters play a crucial role in clustering. Coalescence is the mechanism that most accurately represents the change in size and density for He and vacancy clusters. This occurs primarily through the coalescence of the smallest clusters. So cluster evolution is largely driven by He accumulation and coalescence, while bias plays a minor role. Moreover the generalized vacancy mechanism through Brownian motion of the He-vacancy clusters gives good agreement in the model with experiment.
The Clustering Properties of He and H in Tungsten Studied by Density Functional Theory
Work here focuses on single vacancies and vacancy clusters in tungsten. Ab initio calculations using VASP were made to determine vacancy formation, migration and binding energies. Calculation shows that di-vacancy binding energies are negative, meaning that they are unstable. In fact small vacancy clusters are not stable until they reach a size of at least 4 vacancies, so they are difficult to nucleate but become stable at larger size. This appears to be in agreement with some experimental work.
Very large binding energies for He-He pairs were found, even without the presence of vacancies, when the He atoms were on tetrahedral sites. Calculations have shown that He has strong binding not only to vacancies or impurities, but also to other He atoms, which helps explain why it is difficult to determine the most favorable site for a single He atom, because it is not likely to be free and unbound to some other atom or defect type. One difference between H and He is that while He will move into a vacancy site, H is more likely to move into an octahedral interstitial site. He binding energies with impurity atoms are always higher than H, so it is much easier to eject a H atom from a cluster than to eject the He atom.
Microstructures of Stainless Steels Irradiated in Fast and Thermal Neutron Spectra and with Protons
Experimental data of irradiated microstructure of stainless steel comes from fast neutron, thermal neutron and proton irradiations. The proton irradiations were compared with neutron irradiation data to ensure that comparable effects on microstructure are observed, whereby proton irradiations could then be used to complete the matrix of experimental alloy data. Objectives are to compare RIS, loop microstructure, and hardening in the alloys among different approaches for irradiation.
RIS results show very similar levels of Cr depletion and Ni enrichment among all of the alloys, regardless of irradiation particle type. Radiation-induced hardening data is similar, where, while some differences in magnitude exist between alloys, the hardening data for the same alloy is similar among particle types. Loop microstructure does not compare as favorably, however. Proton irradiation generally has a lower loop density and larger loop size than neutron irradiation.
Using the dispersed-barrier hardening model, the radiation-induced hardening can be calculated based on dislocation loop microstructure and compared to experimentally measured hardening. Comparisons show that for proton irradiation, calculated vs. measured compares well; the same is not as true for neutron irradiation. The implication is that large loops may affect hardening differently from smaller loops. If the smaller loops are ignored when calculating hardening with the dispersed-barrier hardening model, then suddenly calculation vs. experiment matches more closely, with the highest correlation reached when loops of size below 3 nm are neglected in calculations.
The result is that either small dislocation loops are not affecting hardening like the large loops or else the dispersed-barrier hardening model does not accurately correlate loop density and size to hardening due to the fact that model development occurred prior to instrumentation that permitted observation of small loops which can today be imaged.
IASCC Initiation Due to Localized Deformation in Austenitic Stainless Steels
Since neither RIS nor hardening alone is the sole cause for IASCC, so there must be some other mechanism that also contributes to the initiation of SCC due to irradiation. Some experimental observations of cracked alloys have seen dislocation slip steps that were larger on cracked samples when looking at fracture surfaces. One factor that considers slip bands is the stacking fault energy (SFE), but also irradiation itself. To analyze the effect of SFE, alloys were chosen with a wide spread in SFE values, and then the alloys are irradiated. The irradiated microstructure shows smaller loop size and larger loop density as the alloy SFE decreases. In other words, the alloy with highest SFE had the lowest loop density but the highest loop size.
In cracking tests, slip lines were inhomogeneous with low strain but became more homogeneous with increasing strain of the sample. As the SFE increased, the strain-to-failure for the alloy increases. In other words, an alloy with SFE required much higher strain prior to failure. Furthermore, higher SFE had much less cracking.
As slip lines intersect the grain boundaries, it can result in localized grain boundary sliding. Observation of the slip channels provides evidence that the slip lines can initiate cracking when they intersect the grain boundary. Moreover, far more localized deformation is seen in alloys with low SFE, which helps to explain the higher amount of cracking and the lower strain-to-failure for low-SFE alloys.
Austenite Reversion in High Cr Ferritic/Martensitic Steels by High Fluence Neutron Irradiation
FV448 is a 12CrMoVNb F/M stainless steel. Two different types of this alloy were irradiated with neutrons, one being a parent plate with little Ni and a weld material containing nearly 5% Ni. In the parent plate, the lath structure was retained after irradiation, and no austenite was formed. However, in the weld metal, neutron irradiation caused partial transformation of martensite into a combination austenite/ferrite matrix. While the ferrite was free of voids, the austenite contained a large amount of void swelling. The mechanism for the transformation was determined to be the solute segregation of Ni, since Ni is a strong austenite former and Ni enrichment to austenite regions was observed.
Thermodynamic modeling was performed of the alloy compositions to determine Gibbs energy of the alloy phases and determine stability of the phases before and after irradiation. The energy capture of dislocation loop formation due to neutron irradiation is fed into the thermodynamic software to determine how the energy input can transform the martensite in ferrite and austenite. The transformation is a diffusion-free transformation and the critical austenite transformation temperature is high enough in the weld metal to allow austenite to form. Meanwhile the parent plate has an austenite transformation temperature that is too low, and hence no formation of austenite is observed.
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