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Terentyev D.A.,Belgian Institute for Nuclear Sciences | Osetsky Yu.N.,Oak Ridge National Laboratory | Bacon D.J.,University of Liverpool
Acta Materialia | Year: 2010

Dislocation segments with Burgers vector b = 〈1 0 0〉 are formed during deformation of body-centred-cubic (bcc) metals by the interaction between dislocations with b = 1/2〈1 1 1〉. Such segments are also created by reactions between dislocations and dislocation loops in irradiated bcc metals. The obstacle resistance produced by these segments on gliding dislocations is controlled by their mobility, which is determined in turn by the atomic structure of their cores. The core structure of a straight 〈1 0 0〉 edge dislocation is investigated here by atomic-scale computer simulation for α-iron using three different interatomic potentials. At low temperature the dislocation has a non-planar core consisting of two 1/2〈1 1 1〉 fractional dislocations with atomic disregistry spread on planes inclined to the main glide plane. Increasing temperature modifies this core structure and so reduces the critical applied shear stress for glide of the 〈1 0 0〉 dislocation. It is concluded that the response of the 〈1 0 0〉 edge dislocation to temperature or applied stress determines specific reaction pathways occurring between a moving dislocation and 1/2〈1 1 1〉 dislocation loops. The implications of this for plastic flow in unirradiated and irradiated ferritic materials are discussed and demonstrated by examples. © 2009 Acta Materialia Inc.

Leenaers A.,Belgian Institute for Nuclear Sciences | Van Den Berghe S.,Belgian Institute for Nuclear Sciences | Detavernier C.,Ghent University
Journal of Nuclear Materials | Year: 2013

Recent attempts to qualify the LEU(Mo) dispersion plate fuel with Si addition to the Al matrix up to high power and burn-up have not yet been successful due to unacceptable fuel plate swelling at a local burn-up above 60% 235U. The root cause of the failures is clearly related directly to the formation of the U(Mo)-Al(Si) interaction layer. Excessive formation of these layers around the fuel kernels severely weakens the local mechanical integrity and eventually leads to pillowing of the plate. In 2008, SCK·CEN has launched the SELENIUM U(Mo) dispersion fuel development project in an attempt to find an alternative way to reduce the interaction between U(Mo) fuel kernels and the Al matrix to a significantly low level: by applying a coating on the U(Mo) kernels. Two fuel plates containing 8gU/cc U(Mo) coated with respectively 600 nm Si and 1000 nm ZrN in a pure Al matrix were manufactured. These plates were irradiated in the BR2 reactor up to a maximum heat flux of 470 W/cm2 until a maximum local burn-up of approximately 70% 235U (∼50% plate average) was reached. Awaiting the PIE results, the advantages of applying a coating are discussed in this paper through annealing experiments and TRIM (the Transport of Ions in Matter) calculations. © 2013 Elsevier B.V. All rights reserved.

Jansson V.,Belgian Institute for Nuclear Sciences | Jansson V.,University of Helsinki | Malerba L.,Belgian Institute for Nuclear Sciences
Journal of Nuclear Materials | Year: 2013

Neutron irradiation induces in steels nanostructural changes, which are at the origin of the mechanical degradation that these materials experience during operation in nuclear power plants. Some of these effects can be studied by using as model alloy the iron-carbon system. The Object Kinetic Monte Carlo technique has proven capable of simulating in a realistic and quantitatively reliable way a whole irradiation process. We have developed a model for simulating Fe-C systems using a physical description of the properties of vacancy and self-interstitial atom (SIA) clusters, based on a selection of the latest data from atomistic studies and other available experimental and theoretical work from the literature. Based on these data, the effect of carbon on radiation defect evolution has been largely understood in terms of formation of immobile complexes with vacancies that in turn act as traps for SIA clusters. It is found that this effect can be introduced using generic traps for SIA and vacancy clusters, with a binding energy that depends on the size of the clusters, also chosen on the basis on previously performed atomistic studies. The model proved suitable to reproduce the results of low (<350 K) temperature neutron irradiation experiments, as well as the corresponding post-irradiation annealing up to 700 K, in terms of defect cluster densities and size distribution, when compared to available experimental data from the literature. The use of traps proved instrumental for our model. © 2013 Elsevier B.V. All rights reserved.

Terentyev D.,Belgian Institute for Nuclear Sciences | Osetsky Yu.N.,Oak Ridge National Laboratory | Bacon D.J.,University of Liverpool
Scripta Materialia | Year: 2010

Molecular dynamics simulation was used to investigate reactions of a frac(1, 2) 〈 1 1 1 〉 {1 1 0} edge dislocation with interstitial dislocation loops of frac(1, 2) 〈 1 1 1 〉 and 〈1 0 0〉 type in a model of iron. Whether loops are strong or weak obstacles depends not only on loop size and type, but also on temperature and dislocation velocity. These parameters determine whether a loop is absorbed on the dislocation or left behind as it glides away. Absorption requires glide of a reaction segment over the loop surface and cross-slip of dipole dislocation arms attached to the ends of the segment: these mechanisms depend on temperature and strain rate, as discussed here. © 2010 Acta Materialia Inc.

Xu H.,Oak Ridge National Laboratory | Stoller R.E.,Oak Ridge National Laboratory | Osetsky Y.N.,Oak Ridge National Laboratory | Terentyev D.,Belgian Institute for Nuclear Sciences
Physical Review Letters | Year: 2013

The interstitial loop is a unique signature of radiation damage in structural materials for nuclear and other advanced energy systems. Unlike other bcc metals, two types of interstitial loops, 1/ 2âŸ̈111⟩ and 100, are formed in bcc iron and its alloys. However, the mechanism by which 100 interstitial dislocation loops are formed has remained undetermined since they were first observed more than fifty years ago. We describe our atomistic simulations that have provided the first direct observation of 100 loop formation. The process was initially observed using our self-evolving atomistic kinetic Monte Carlo method, and subsequently confirmed using molecular dynamics simulations. Formation of 100 loops involves a distinctly atomistic interaction between two 1/2âŸ̈111⟩ loops, and does not follow the conventional assumption of dislocation theory, which is Burgers vector conservation between the reactants and the product. The process observed is different from all previously proposed mechanisms. Thus, our observations might provide a direct link between experiments and simulations and new insights into defect formation that may provide a basis to increase the radiation resistance of these strategic materials. © 2013 American Physical Society.

Pshirkov M.S.,RAS Institute for Nuclear Research | Tinyakov P.G.,Belgian Institute for Nuclear Sciences | Urban F.R.,Free University of Colombia
Monthly Notices of the Royal Astronomical Society | Year: 2013

We study the influence of the random part of the galactic magnetic field on the propagation of ultrahigh energy cosmic rays. Within very mild approximations about the properties of the electron density fluctuations in the Galaxy, we are able to derive a clear and direct relation between the observed variance of rotation measures and the predicted cosmic ray deflections. Remarkably, this is obtained bypassing entirely the detailed knowledge of the magnetic properties of the turbulent plasma. Depending on the parameters of the electron density spectrum, we can either directly estimate the expected deflection, or constrain it from above. Thanks to the latest observational data on rotation measures, we build a directiondependent map of such deflections. We find that over most of the sky the random deflections of 40 EeV protons do not exceed 1°-2°, and can be as large as 5? close to the Galactic plane. © 2013 The Authors Published by Oxford University Press on behalf of the Royal Astronomical Society.

De Bremaecker A.,Belgian Institute for Nuclear Sciences
Journal of Nuclear Materials | Year: 2012

In the 1960s in the frame of the sodium-cooled fast breeders, SCK•CEN decided to develop claddings made with ferritic stainless materials because of their specific properties, namely a higher thermal conductivity, a lower thermal expansion, a lower tendency to He-embrittlement, and a lower swelling than the austenitic stainless steels. To enhance their lower creep resistance at 650-700 °C arose the idea to strengthen the microstructure by oxide dispersions. This was the starting point of an ambitious programme where both the matrix and the dispersions were optimized. A purely ferritic 13 wt% Cr matrix was selected and its mechanical strength was improved through addition of ferritizing elements. Results of tensile and stress-rupture tests showed that Ti and Mo were the most beneficial elements, partly because of the chi-phase precipitation. In 1973 the optimized matrix composition was Fe-13Cr-3.5Ti-2Mo. To reach creep properties similar to those of AISI 316, different dispersions and methods were tested: internal oxidation (that was not conclusive), and the direct mixing of metallic and oxide powders (Al 2O 3, MgO, ZrO 2, TiO 2, ZrSiO 4) followed by pressing, sintering, and extrusion. The compression and extrusion parameters were determined: extrusion as hollow at 1050 °C, solution annealing at 1050 °C/15 min, cleaning, cold drawing to the final dimensions with intermediate annealings at 1050 °C, final annealing at 1050 °C, straightening and final aging at 800 °C. The choice of titania and yttria powders and their concentrations were finalized on the basis of their out-of-pile and in-pile creep and tensile strength. As soon as a resistance butt welding machine was developed and installed in a glove-box, fuel segments with PuO 2 were loaded in the Belgian MTR BR2. The fabrication parameters were continuously optimized: milling and beating, lubrication, cold drawing (partial and final reduction rates, temperature, duration, atmosphere and furnace). Specific non-destructive tests (ultrasonic and eddy currents) were also developed. In-pile creep in argon and in liquid sodium was deeply studied on pressurized segments irradiated up to 75 dpa NRT. Finally two fuel assemblies cladded with such ODS alloys were irradiated in Phenix to the max dose of 90 dpa. Creep deformation and swelling were limited but the irradiation-induced embrittlement became acute. The programme was stopped shortly after the Chernobyl disaster, before the embrittlement problem was solved. © 2011 Elsevier B.V. All rights reserved.

Terentyev D.,Belgian Institute for Nuclear Sciences | Bergner F.,Helmholtz Center Dresden | Osetsky Y.,Oak Ridge National Laboratory
Acta Materialia | Year: 2013

The effect of chromium on iron hardening via segregation on dislocation loops was studied by atomic scale computer modeling. A combination of Monte Carlo and molecular dynamics techniques together with the recently determined Fe-Cr interatomic potentials fitted to ab initio data was used to investigate Cr segregation on 1/2〈1 1 1〉 interstitial dislocation loops and its impact on the interaction with moving dislocations. The Monte Carlo results reveal that Cr atoms segregate to the loop tensile strain region and dissolve well above the temperature corresponding to the solubility limit. The molecular dynamics results demonstrated that local micro-chemical changes near the loop reduce its mobility and increase the strength. The stress to move a dislocation through the array of Cr "decorated" loops increases due to modification of the dislocation-loop interaction mechanism. A possible explanation for a number of experimental observations being dependent on the radiation dose and for Cr concentration effects on the yield stress is given on the basis of the modeling results. © 2012 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

Terentyev D.,Belgian Institute for Nuclear Sciences | Monnet G.,Électricité de France | Grigorev P.,Belgian Institute for Nuclear Sciences
Scripta Materialia | Year: 2013

We propose a computationally fast and physically justifiable method to treat dislocation loops as stochastic thermally activated finite-size obstacles in discrete dislocation dynamics simulations. The method was parameterized using molecular dynamics data for the interaction of dislocations with a 0/2〈1 1 1 dislocation loops. As demonstration, the method is applied to rationalize experimental hardening of neutron-irradiated iron. The obtained results show good agreement with experimental data. © 2013 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

Terentyev D.,Belgian Institute for Nuclear Sciences | Martin-Bragado I.,IMDEA Madrid Institute for Advanced Studies
Scripta Materialia | Year: 2015

The impact of carbon content on the evolution of dislocation loops in iron-carbon was studied by object kinetic Monte Carlo, explicitly introducing carbon atoms and their atomic features for the first time. We demonstrate that the saturated loop density strongly depends on carbon content and temperature, in good agreement with in situ irradiation microscopy studies. The physical processes responsible for the accumulation and long-range migration of the loops are rationalized with implications for nanostructural evolution in commercial steels upon low-dose-rate neutron irradiation. © 2014 Acta Materialia Inc.

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