For commercial zirconium alloy cladding used in light water reactor (LWR), hydrogen pickup increases with the extent of waterside corrosion, thereby causing cladding ductility to decrease. In addition, pre-storage drying-transfer operations might expose cladding to higher temperatures and higher pressure, which in turn will introduce higher tensile hoop stresses relative to normal operation in-reactor and pool storage. Under these conditions, hydrides could be redistributed and provide an additional embrittlement mechanism. In order to better understand fuel behavior, it is important for safety analyses to evaluate mechanical properties induced by hydrogen charging. As a means of simulating the used fuel behavior, hydrided Zr samples were fabricated at Oak Ridge National Laboratory (ORNL). Hydrided Zircaloy-4 samples were produced by a gas charging method to levels that encompass the range of hydrogen concentrations observed in current used fuel. For low hydrogen content samples, the hydrided platelets appear elongated and needle-like, orientated in the circumferential direction. Mechanical testing was carried out by the ring compression method at various temperatures. Samples with higher hydrogen concentration exhibited in lower strain before fracture and reduced maximum load. The trend between temperature and ductility was also very clear: increasing temperatures resulted in increased ductility of the hydrided cladding. In this paper we present the results of neutron diffraction studies to determine the relationship of the changes in ductility to the ratios of various phases of the zirconium hydride as a function of temperature from ambient to 400C.

Hydrogen pickup in zirconium alloys is a prominent challenge in front of the design of these alloys for fuel cladding in nuclear reactors. In 1960 a volcano-like dependence of the hydrogen pickup fraction was identified across the 3d transition metals that are used to alloy zirconium [1]. This empirical observation was used subsequently in the design of zirconium alloys without a physical understanding of its origin. Here we show using a combination of density functional theory calculations and thermodynamic analysis that hydrogen solubility in ZrO₂ - The native passive layer that grows on zirconium alloys- exhibits a similar volcano-like dependence on the 3d transition metals. We found that the origin of this volcano is the variation in the ability of the 3d transition metals to p-type dope ZrO₂. This provide a physical understanding for the experimental results.
Recasting the calculated hydrogen solubility in ZrO₂ on the electron chemical potential space gives rise to a valley-like dependence. For designing zirconium alloys resistant against hydrogen pickup, we suggest targeting either a dopant that thermodynamically minimizes the solubility of hydrogen in ZrO₂ at the bottom of this valley, or a dopant that maximizes the electron chemical potential and kinetically accelerates hydrogen reduction and H₂ evolution at the surface of ZrO₂.
The paradigm we present here for zirconium alloys opens the door to a general understanding for the role of the native oxide passive layer in mitigating the ingress of hydrogen into other alloy systems.
[1] B. Cox, M. J. Davies, A. D. Dent, “The oxidation and corrosion of zirconium and its alloys. Part X. Hydrogen absorption during oxidation in steam and aqueous solutions.”, AERE-M621, HARWELL, 1960.

Zirconium alloy of composition Zr-2.5Nb is frequently used as a fuel cell cladding and structural material in nuclear reactors, due to its intrinsic neutron transparency and resistance to radiation damage. The mechanical properties and formability of this material depend strongly on an engineered microstructure, which is obtained through well-designed thermo-mechanical processing routes. Therefore, it is important to know the transformation and defect kinetics and their mechanisms.
We have performed in-situ neutron diffraction experiments while the specimens undergo heating and cooling cycles. Quantitative phase analysis delivers the overall phase composition, while changes in lattice parameter allow to conclude of the momentary composition, as phase transformation and segregation effects occur. Furthermore, the effect of primary extinction of neutron radiation allows to follow the defect kinetics in the high-temperature beta phase. Thus recovery through annihilation of dislocation leads to more and more perfect crystal grains. On cooling, precipitating alpha phase distorts the lattice of the beta grains in a nucleation-and-growth process. Plastic deformation has been applied during the experiments in order to introduce a source of dislocations and validate the observed effects.
The presented in-situ methods give valuable insights to the material in real time and can be applied to a wide range of processing routes and materials.

Fuel cladding tubes are exposed to high temperature and pressure in nuclear reactors and despite the importance of cladding corrosion at normal and accidental conditions, there is a lack of understanding of the reaction mechanism at the solid-fluid interfaces [1].
The study focuses on designing and building a portable sample environment suitable for in situ investigation of interfacial interactions at high pressure and temperature conditions. The reaction cell has been optimized for in situ synchrotron techniques. X-ray fluorescence and X-ray diffraction will be used to elucidate the corrosion of the conventional and proposed cladding alloys in a reactor-like environment. The cell design was optimized for submicron resolution X-ray spectroscopy and X-ray powder diffraction first light beamlines in the national synchrotron light source (NSLS-II) at Brookhaven national laboratory. The cell will be available to all users through the competitive user facility proposal process.
The cell design was adapted from J. Diefenbacher, et. al., [2]. However, it was modified for corrosion applications. The reaction cell and sample holder were made from hastelloy due to its corrosion resistant. The core has a cylindrical diameter of 5 mm with grooves to support the sample holder in the center to assure uniform oxidation on both sides of the sample. Sample will be exposed to saturated steam at temperatures and pressures up to 400°C and 1500 psi. The steam temperature will be recorded with a thermocouple imbedded in the reaction cell. The ratio between the sample surface area and the cell volume was designed to be 0.1 m²/L, to comply with ASTM standards for corrosion tests of zirconium alloys [3]. The core has a 9 mm diameter cut on both sides to place the windows which are separated from the outer frames by aluminum gaskets to reduce stresses. Moissanite windows will be used for XRD measurements, while glassy carbon (GC) will be used for XRF measurements. The reaction cell is surrounded by copper heating block to control the temperature independently from the rest of the system. However, the overall system pressure is controlled with a pressurizer to avoid large fluctuations with temperature change. A lab view platform is used for remote monitoring of pressure and temperature. A distribution manifold is used for controlling the circulation of gases and fluids in the system. The system has built in safety features; 2 burst disks on the pressurizer to avoid over pressure and an alarm system to warn for unexpected changes in pressure and temperature.
[1] A. Yilmazbayhan, et. al., International Conference on Environmental Degradation of Materials, 201 (2005).
[2] J. Diefenbacher, et. al., Review of Scientific Instruments, 76, 015103 (2005).
[3] ASTM standards, G2/G2M-06, 2011. Corrosion Testing of Products of Zirconium, Hafnium, and Their Alloys in Water at 680°F or in Steam at 750°F.

Zirconium alloys are used as cladding materials in nuclear reactors. Due to the large amount of point defects created under irradiation, they experience a dimensional change without applied stress. These defects evolve towards larger clusters, leading to prismatic dislocation loops, both of interstitial and vacancy type. When the irradiation dose increases, a growth enhancement is observed, correlated to the appearance of basal vacancy loops, called loops. Understanding the formation of these clusters is of prime importance to model the kinetic evolution of the microstructure under irradiation. This requires first to know their relative stability, in particular for vacancy clusters for which different types can coexist.
We propose here an atomic scale study of the stability properties of vacancy clusters in hexagonal close-packed Zr (cavities and dislocation loops). Our modeling approach is based both on density functional theory and empirical potentials. Considering the vacancy-vacancy interactions and the stability of small vacancy clusters, we establish how to build the larger clusters. The study of extended vacancy clusters is then performed using continuous laws for defect energetics. Once validated with an empirical potential, these laws are parameterized with ab initio data. Our work shows that the easy formation of loops can be explained by their thermodynamic properties [1].
[1] C. Varvenne, O. Mackain, and E. Clouet, Acta Mater. in press (2014);
http://dx.doi.org/10.1016/j.actamat.2014.06.012
This work has been funded by Areva.

Predictive calculations of high-temperature phenomena in strongly anharmonic crystals and high-temperature phases has long been an outstanding challenge in condensed matter physics, to the detriment of nuclear materials research. In particular, the ZrH_2-x system, which plays an essential role in fuel rod corrosion, has a complex phase diagram that includes a high-temperature cubic phase that is dynamically unstable according to density functional theory. The recently developed anharmonic potential cluster expansion framework resolves this problem by enabling construction of general and arbitrarily accurate atomistic Hamiltonians that are parameterized from first principles. Using Monte Carlo or molecular dynamics simulation, the anharmonic cluster expansion allows us to study martensitic transitions and simulate high-temperature phases that are mechanically unstable at 0K.
We use the anharmonic potential cluster expansion, along with Monte Carlo simulations, to calculate the stress-temperature phase diagram of ZrH₂, which exhibits cubic, tetragonal, and orthorhombic phase stability. We also calculate elastic constants of the high- and low-temperature phases, finding that temperature and stress-state significantly affect the stiffness and elastic anisotropy of ZrH₂, including noticeable critical softening along the second-order transitions.

The martensitic ODS steels have a distinct advantage in processing and manufacturing over full ferritic ODS steels, since the α/γ transformation is reversible reaction with large driving force, compared with recrystallization used in fully ferritic type.
Martensitic 9CrODS steels, 9Cr-0.13C-2W-0.2Ti-0.35Y₂O₃, is a candidate cladding materials for Generation IV fast reactor fuel, and F83H-ODS, 8Cr-0.16C-2W-0.2Ti-VTa-0.40Y₂O₃, is candidate blanket materials of the advanced fusion reactors. The ferrite is superimposed in both materials, although their structures are predicted to be full martensite from a computed phase diagram. As increasing a driving force for α/γ reverse transformation by more addition of carbon as well as decreasing oxide particles pining force by reducing Y₂O₃ content, their structures are modified to the single martensite without ferrite. Hence, the ferrite in 9CrODS and F82H-ODS is a metastable phase and involves extremely finer nano-size oxide particles, which is responsible for significantly improved high-temperature strength in martensitic ODS steels.
Block boundaries inside the tempered martensite serve sites for deformation and softening, thus in terms of severe hot-rolling at the γ-region, the microstructure of the tempered martensite was modified to the transformed-ferrite with coarser grains. This processing doesn&’t follow an established knowledge that is widely accepted as ultrafine grain formation by thermo-mechanical controlled processing. The ferrite transformation from the austenite follows Kurdjumov-Sacks relationship; there are 24 variants in crystalline orientation for the ferrite nucleated from austenite. It was verified that variant selection of transformed ferrite grains can be restricted, and thus neighboring ferrite grains are coalesced and coarsened. It is worth noting that ferrite grain coarsening leads to extremely high tensile and creep rupture strength, keeping excellent ductility at 973 K. Such improvement is attributed to the coarsened ferrite grains that suppress the localized deformation at the martensite block boundaries.

Stacking fault tetrahedra are detrimental defects in neutron or proton irradiated structural metals with face-centered-cubic structures. Their removal is very challenging and typically requires annealing at very high temperatures, incorporation of interstitials or interaction with mobile dislocations. We present an alternative solution to remove stacking fault tetrahedra discovered during room-temperature, in situ Kr ion irradiation of epitaxial nanotwinned Ag with an average twin spacing of ~ 8 nm. A large number of stacking fault tetrahedra are removed during their interactions with abundant coherent twin boundaries [KY Yu et al, Nature Communications, 4 (2013) 1377]. Consequently the density of stacking fault tetrahedra in irradiated nanotwinned Ag is much lower than that in its bulk counterpart. Two fundamental interaction mechanisms are identified, and compared to predictions by molecular dynamics simulations. In situ studies also reveal a new phenomenon: radiation induced frequent migration of coherent and incoherent twin boundaries [KY Yu et al, Scripta Mater, 69 (2013) 385]. Such twin boundary migration is closely correlated to the absorption of radiation generated dislocation loops. Potential migration mechanisms are discussed. This research is funded by NSF-DMR-Metallic Materials and Nanostructures Program.

Nanostructured ferritic alloys (NFAs) have outstanding high tensile and creep strength permitting operation at high temperature and manifest extreme tolerance to radiation damage^1,2. These remarkable properties are due to an ultrahigh density of Ti-Y-O enriched nano-features that provide dispersion strengthening, help stabilize dislocation and fine grain structures, reduce excess concentrations of displacement defects etc. To achieve these properties, NFAs are fabricated by mechanical alloying of metallic and yttria powders. The processing routes, i.e., casting versus mechanical alloying, as well as parameters such as the milling times, are key considerations on the desired microstructure. Atom probe tomography analysis has shown that coarse oxide particles are present after relatively short milling times of 5 and 20 h. Milling for at least 40 h appears to required to produce a uniform distribution of solutes in mechanically-alloyed flakes. After hot extrusion, the microstructure consists of a-Fe, high number densities of Ti-Y-O-vacancy-enriched nanoclusters, and coarser Y₂Ti₂O₇ and Ti(O,C,N) precipitates on the grain boundaries. The as-cast condition has a distinctly different microstructure consisting of a-Fe with 50-100 mm irregularly-shaped Y₂Ti₂O₇ pyrochlore precipitates with smaller embedded precipitates with the Al₅Y₃O₁₂ (yttrium-aluminum garnet) crystal structure. The nano-hardnesses were also found to be substantially different, i.e., 4 and 8 GPa for the as-cast and as-extruded conditions, respectively. These variances can be due to the high vacancy content introduced by mechanical alloying, and the strong binding energy of vacancies with O, Ti, and Y atoms retarding diffusion.
Research sponsored by the Materials Sciences and Engineering Division, Office of Basic Energy Sciences, US Department of Energy. The microscopy was supported through a user project supported by ORNL&’s Center for Nanophase Materials Sciences (CNMS), which is sponsored by the Scientific User Facilities Division, Office of Basic Energy Sciences, U.S. Department of Energy.
1) M. K. Miller, C. M. Parish and Q. Li, Materials Science and Technology 29, 1174 (2013)
2) M. C. Brandes, L. Kovarik, M. K. Miller, M. J. Mills, J Mater Sci 47, 3913 (2012)

Advanced oxide dispersion strengthened (ODS) ferritic alloys have been developed for fusion energy application not only due to the exceptional high-temperature tensile strength and creep resistance, but also for the excellent tolerance to high dose irradiation. By adopting Y₂O₃ powder as an extra alloying addition, stable nanoclusters (NCs) other than oxide precipitates are observed after the mechanical alloying process. For 14YWT [1-6], 2-4nm diameter size of Y-Ti-O enriched NCs were observed to be extremely stable up to 0.92 of the melting temperature. In ODS-Eurofer, Y-V-O enriched NCs were observed with the size of 1-5nm diameter or as a 1-3nm thick shell out of the Y₂O₃ precipitates. Different from the conventional nano-phase material which is metastable in nature, these NCs do not coarsen at the elevated temperatures. Thus it is important to understand the formation and stability mechanism of the unique material state of these NCs as well as the role of Yttrium atoms during the formation process.
We developed an internal strain induced formation (ISIF) model to study the formation and stability of Y-Ti-O enriched nanoclusters in 14YWT. In this presentation, we will adopt this ISIF model to study the formation and stability of Y-V-O enriched nanoclusters in Eurofer97. The essential stability condition of these nanoclusters is the exceptionally low interfacial energy compared with the interior energy of these nanoclusters. The strain energy accumulated within the nanoclusters is mainly from the solute-solute repulsive interaction and the Y-O:vacancy interaction due to the un-relaxed Y atom at the lattice site. The Y-V-O-enriched nanoclusters are predicated to form in at a lower O concentration range. The size of Y-V-O-enriched nanoclusters is closely related with the Y/V composition ratio.

The Zr-based alloys used for nuclear fuel cladding suffer corrosion and hydriding in-reactor, which degrade material properties. It is well known that the corrosion rate and stability of the protective oxide layer are strong functions of the alloying composition which result in different but quite reproducible pre transition corrosion kinetics between alloys. Recent results from synchrotron based x-ray near edge absorption spectroscopy, transmission electron microscopy and cold neutron prompt gamma activation analysis of autoclaved samples show that hydrogen pickup fraction is equally affected, varying during the corrosion process and from alloy to alloy. The results can be understood in terms of a couple current charge compensation model recently developed, according to which the hydrogen pickup increases as the oxide electronic conductivity decreases. These efforts will be reviewed in his talk.

In order for advanced nuclear reactor concepts to come to fruition, a high level of social and scientific confidence must be achieved. Predictive physics-based modeling has emerged as a path to increase at least the scientific confidence of various proposed materials in the extreme environments that are associated with the next generation of nuclear reactors. For these models to be physics-based, the underlying mechanisms governing the response of both current and proposed materials in these combinations of environments must be understood down to the nanometer scale. In-situ transmission electron microscopy (TEM) experiments provide the ideal test chamber for many of the experiments needed to elucidate the active mechanisms.
The extensive work done over the past several decades to understand materials response to extreme environments and combinations of extreme environments will be reviewed in this talk. These combinations included radiation, mechanical loading, gas environments, and liquid metal exposure.
The last half of the talk will focus on the recent advancements made at Sandia National Laboratories&’ in-situ ion irradiation TEM facility in advancing this field of study through further combination of in-situ TEM capabilities. This facility has demonstrated the capability to do in-situ ion irradiation with a wealth of high energy heavy ions, while concurrently implanting the TEM sample with helium and deuterium. In addition, the high contrast pole piece of this microscope permits up to 162#730; of alpha tilt for thorough sequential tomographic series during in-situ TEM experiments. This presentation will also highlight the ongoing work in combining ion irradiation with quantitative mechanical testing, microfluidic cells, gas-heating environments. This presentation will conclude with speculation on how recent inclusion of large data processing, microfabrication, object tracking, and single photon detection will affect the evolution of in-situ TEM. In summary, this talk will highlight the recent advancements in combining various types of in-situ TEM experiments in an effort to better understand the basic mechanisms governing the structural evolution in materials subject to the harsh environments expected in advanced nuclear reactors.
This work is partially supported by the Division of Materials Science and Engineering, Office of Basic Energy Sciences, U.S. Department of Energy. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy&’s National Nuclear Security Administration under contract DE-AC04-94AL85000.

The void nucleation behavior in hcp metals is studied by combining atomistic simulations and in-situ high-resolution TEM. Under electron irradiation in Mg, the voids are observed to take polyhedron shapes in the nucleation stage, with the facets being low-energy surfaces including {0001} and {01i1}. Due the competition between thermodynamic and kinetic aspects, with increasing size the voids experience a transition in shapes from elongated platelet to a nearly equiaxial geometry. Molecular dynamics simulations show that the inter-facet vacancy diffusion between {0001} and {01i1} surfaces is anisotropic and is dependent on the thickness of the voids along the <0001> direction. When the thickness is small, it is easier for a vacancy to diffuse from the side {01i1} to the basal {0001} surface, and in-situ TEM show that the voids quickly thicken to a few layers. However, beyond a certain thickness of about 5 layers, the favorable inter-facet vacancy diffusion is revered, and the voids stop thickening along <0001> and grow along the length direction in the basal plane, leading to an elongated platelet shape. During further growth, nucleation of new vacant layers at the center of {0001} surfaces becomes possible due to the large {0001} surface area, and the voids transition into a nearly equiaxial geometry given by the Wulff construction. This geometry minimizes the total surface energy of the voids and is therefore favored thermodynamically at large sizes.

The next generation nuclear reactors require materials that can not only serve at elevated temperatures but also resist irradiation damage under high neutron doses. Recent findings show that it is nano-sized oxides that lead to high creep strength and good irradiation damage tolerance for high temperature operations in ferritic oxide dispersion strengthened (ODS) alloys. Most of nano-sized oxides in current ODS alloys are Y based refined oxides with Ti, Al, Ta or Zr additions. However, little is known whether there exist other types of oxides that have similar ultrafine sizes and dispersions in the matrix. Here we report the formation of high density of ZrO₂ nano particles in Fe-14Cr alloy powders with Zr addition synthesized by mechanical alloying. The nano ZrO₂ particles were found uniformly dispersed in the ferritic matrix with an average size less than 5 nm, which stabilize the nanocrystalline matrix after annealing at 9000C for 1h. These oxides are carefully characterized by means of EDS elemental mapping in STEM, HRTEM and electron diffraction. The thermal stability of the nanocrystalline ferritic matrix alloy powders is largely attributed to the Zener pinning of grain boundaries by the nano-sized, highly dispersed ZrO₂ particles. More importantly, the size and dispersion of the ZrO₂ particles are comparable to those of Y-Ti-O enriched oxides reported in irradiation-resistant ODS alloys. Our findings suggest other type of oxides with ultrafine particle sizes and dispersions in the ferritic matrix, similar to Y-based oxides in ODS alloys i.e. a possible new irradiation resistant material for nuclear energy applications. The improved high-temperature nano grain size stabilization by these other nanoscale oxides can contribute to additional resistance to radiation damage.

Metal-oxide multilayer nanocomposite was used as a model for understanding the evolution of interface structure and diffusion behavior of Cr in irradiated oxide-dispersion-strengthened (ODS) steels. The thin films, containing chemical sharp metal-oxide interfaces (FeCr-TiO₂, FeCr-Y₂O₃), were deposited on MgO (100) substrate at 500 °C, ensued with irradiation by 10Mev Ni³⁺ ions at temperature 500 °C. In comparison, the pristine nanocomposite has been annealed at 500 °C. Microchemistry and microstructure evolution of metal/oxide multilayer was investigated by using high resolution transmission electron microscopy, energy-dispersive X-ray spectroscopy and electron energy loss spectroscopy. We found for FeCr-TiO₂ and FeCr-MgO interfaces, radiation enhanced/accelerated Cr diffusion into oxide. But for FeCr-Y₂O₃ interface, radiation cause Cr diffusion into oxide. Meanwhile, amorphization has been enhanced by Cr diffusion. We believe the knowledge obtained from this work provides guidelines for designing metal-oxide composite with desired radiation tolerance under irradiation.

Accurate understandings of structures and behaviors of radiation-produced lattice defects are required for predicting processes of nuclear-fission and fusion materials. Most elementary defects among various types of defects are atomic-size point defects (self-interstitial atoms (SIAs) and vacancies). The most hopeful experimental method for directly detecting behaviors of defects within materials is transmission electron microscopy (TEM) [1-3]. However, even using cutting edge TEM, directly tracing the behaviors of rapidly migrating individual SIAs within comparatively thick specimens is impossible.
Under high-energy electron irradiation, isolated SIAs and vacancies are produced almost spatially homogeneously; therefore, the mesoscopic process of clustering of SIAs, which can be directly observed by high-voltage electron microscopy (HVEM), is expected to reflect SIA behaviors. Here, we tried to extract parameters related to SIA behaviors from the formation process of SIA clusters in the form of dislocation loops, using HVEM. In the present talk, our recent studies e.g. a straightforward estimation of the activation energy for the migration of SIAs in high-purity tungsten [4], are presented.
[1] Arakawa, K. et al., “Changes in the Burgers Vector of Perfect Dislocation Loops without Contact with the External Dislocations,” Phys. Rev. Lett., 96 (2006) 125506.
[2] Arakawa, K. et al., “Observation of the One-Dimensional Diffusion of Nanometer-Sized Dislocation Loops,” Science, 318 (2007) 956.
[3] Arakawa, K., Amino, T., and Mori, H., “Direct Observation of the Coalescence Process between Nanoscale Dislocation Loops with Different Burgers Vectors,” Acta Mater., 59 (2011) 141.
[4] Amino, T., Arakawa, K., and Mori, H., “Activation Energy for Long-Range Migration of Self-Interstitial Atoms in Tungsten Obtained by Direct Measurement of Radiation-Induced Point-Defect Clusters,” Philos. Mag. Lett., 91 (2011) 86.

Small scale mechanical testing offers a wide range of benefits to the nuclear materials community including the volume reduction of activated materials, localized mechanical evaluation, increased statistics on a specimen, making mechanical data accessible to ion beam irradiated materials and separate effects evaluation. In this presentation previous work on nanoindentation and micro compression testing on irradiated materials will be discussed on conventional materials like 304SS as well as more advanced materials like ODS alloys and F/M steels. The importance of size and scaling effects utilizing these techniques will be highlighted on materials irradiated with ion beams and reactors. New methods featuring “lift out” mechanical specimen geometries and testing will be introduced which allows evaluating highly activated samples after reactor irradiations and easy handling. In addition mechanical testing at operation condition (temperature) will be discussed and recent results obtained on 304SS will be compared to literature values. Microstructural characterization in combination with the mechanical properties measured allows correlating the radiation induced defects with the performance degradation of the material in question and therefore is an integrated component of small scale mechanical testing.

FeCr ferritic/martensitic steels are foreseen as strong candidates for future fission and fusion nuclear reactor since they have superior properties under irradiation compared to traditional steels. One of the major concerns is radiation-induced solute redistribution (RISR) as it might change not only the mechanical properties but also the response of the material to corrosive environments. We present a model to study RISR in FeCr alloys that reproduces the thermodynamic and kinetic properties of the system. Using a kinetic Monte Carlo algorithm we are able to study the microstructure evolution of the material under light ion bombardment in the presence of planar perfect sinks, mimicking the effect of grain boundaries. We observe that for a window of temperatures and compositions varying the distance between sinks leads to the mitigation of the Cr redistribution. Therefore, material nano-engineering with the appropriate interfaces could help in reducing the deleterious effects of RISR.

Carbon Nanotubes exhibit exceptional properties in terms of high strength-to-weight, high electrical conductivity, and high thermal conductivity, and have been employed as a reinforcement in various composites and other materials. Their tolerance to radiation environments may be suggested by their response to energetic ion bombardment. We discuss the effects of argon ion bombardment of both thin and thick multiwall carbon nanotube films over a range of 4 to 11 keV at fluence levels up to the order of 10²¹ ions/cm². While individual carbon atoms are readily displaced from a carbon nanotube by bombardment at these energies, these nanotubes also exhibit a self-healing capability. At moderate energies and fluence, if two or more carbon nanotubes are touching and an ion strikes this point, they heal together where a junction or cross-link between them is created and the nanotubes interpenetrate. Even though some of the properties of the carbon nanotubes may be degraded by ion bombardment at non-junction regions, we have demonstrated a bulk cross-linked thin film of randomly oriented multiwall carbon nanotubes with an isotropic thermal conductivity of 2150 W/m-K. At higher energies and fluence, the carbon nanotubes appear to collapse and reform aligned parallel to the incoming ion bombardment trajectory, producing high aspect ratio tapered structures. These structures are, in general, fully dense, unlike the loosely packed random carbon nanotube array from which they originated. There is also a sharp transition at the base of these structures from the dense form to the loose-packed form, suggesting that these structures may inhibit further penetration of the energetic ions.

According to the modern theory the evolution of radiation damage depends on the fraction of interstitial atoms produced in the form of one-dimensionally glissile clusters. These one-dimensionally (1-D) glissile clusters have a low interaction cross-section with other defects and die mainly on grain boundaries creating so-called production bias. In this paper we report the results of an extensive multi-technique atomistic modeling of interstitial clusters mobility in bcc Fe-Ni alloys with Ni content from 0.8 to 10 at.%. We have considered Ni for the Fe-Ni interatomic potential well reproduces a number of related properties including those that were not used in the fitting procedure. We have found that Ni interacts strongly with the edge dislocation on the periphery of interstitial clusters/dislocation loops. The breaking effect is therefore defined by the number of Ni atoms interacting with the cluster at the same time. We have found that the breaking is significant even in low-Ni alloys: for example cluster of 37SIA is practically immobile at <500K in Fe-0.8at.% Ni alloy. Increasing cluster size and Ni content makes leads up to complete immobilization. This has quite broad consequences: 1) increase matrix damage for they now can accumulate in the bulk; 2) the above reduces “production bias” and, therefore, radiation swelling via increasing recombination with vacancies and 3) increases radiation induced hardening for contribute to pin dislocations during deformation. The results obtained help in predicting swelling, microstructure evolution and hardening in Fe-Ni ferritic alloys under irradiation.

Under irradiation, a large amount of vacancies (V) are produced. They strongly interact with interstitial solutes (X) such as carbon (C), nitrogen (N) and oxygen (O) atoms, which are always present in steels, either as alloying elements or as impurities. The V-X attraction influences the mobility of both the solutes and the vacancies. On one hand, a decrease of the vacancy mobility has been revealed experimentally in the presence of carbon and nitrogen, most likely due to the trapping of vacancies at small vacancy-solute complexes [1, 2]. On the other hand, however, it is not clear whether vacancies always reduce the mobility of the interstitial elements.
Density Functional Theory (DFT) calculations have been performed to study the energetic and kinetic properties of V_nX_m clusters. Low-energy configurations of small V_nX_m have been determined. It has been revealed that vacancies enhance the clustering of solutes. Moreover, a systematic comparison of C, N and O - neighbors in the Periodic Table - shows different behaviors of the solutes in the neighborhood of vacancies as a function of the electronic band filling. For instance C atoms tend to decorate the surface of V clusters whereas O atoms will preferentially gather inside the V clusters.
The mobility of the V_nX_m clusters has been carefully studied. We especially focused on the V_nX clusters as it has been shown that V₂ and V₃ are even more mobile than a monovacancy in α-Fe [3]. As a result, all the V₃X have been found to be very mobile. In particular, some clusters can be as mobile as the isolated solutes. Therefore, vacancies may be efficient to drag the interstitial solutes towards sinks such as grain boundaries, dislocations and free surfaces. Also, the result found on the mobility of small V_nN clusters may explain the apparent discrepancy between the resistivity recovery experiments and the DFT data [2]. The interpretation of such experiments may be worth revisiting in the light of the present DFT prediction.
The obtained DFT data have been used to parameterize a Cluster Dynamics model, based on the Rate Theory, which allows to predict the time evolution of the clusters concentration. The consequences of small highly mobile clusters on the kinetic properties of vacancies and solutes under various irradiation conditions have been explored using this model.
[1] S. Takaki et al., Rad. Eff. 79, 87 (1983).
[2] A. L. Nikolaev et al., J. Nucl. Mater. 414, 374 (2011).
[3] C.-C. Fu et al., Nature Mater. 4, 68 (2005).

Irradiation creep leads to significant deformations of reactor structural materials and, for this reason, has been studied already for more than a half of a century. It has been found that in stainless steels it consists of two contributions, one of which depends on swelling rate and the other does not. It is commonly accepted that the former part is described fairly well by the Gittus model, also known as I-creep or dislocation glide-assisted climb model, whereas the origin of latter part is yet unclear. The SIPA (stress-induced preferential absorption) mechanism is often assumed for this part, but we argue that it is not applicable in the case of neutron irradiation because in this case the primary damage is not point defects only, as assumed in the SIPA model. Similar arguments are applicable to other creep mechanisms proposed to date. The swelling-independent creep is even more important for Zr -based alloys, where the dislocation climb is due to radiation growth with much smaller strain rates than those of swelling in steels. In this paper we study the swelling/growth-independent creep and show that it may be explained by taking into consideration the true nature of the primary damage produced by neutrons.

<111> crowdions are the most stable self-interstitial atomic (SIA) defect in the majority of the bcc transition metals, and are formed in large quantities when these metals, and their alloys, are irradiated [1]. In pure metals, they cluster together to form nanometre-scale prismatic loops that, together with their vacancy-formed counterparts, play an important role in irradiation hardening and embrittlement [2]. It is well known that alloys often fare better than pure metals where radiation damage is concerned, and this can probably be attributed to the interactions of vacancies and interstitials with solute atoms (in dilute alloys) and precipitates (in concentrated ones). DFT calculations have determined the binding energies for crowdions and solute atoms in various systems [3-5], and it is important to note that whether the interaction is attractive or repulsive, it will impede the agglomeration of the SIAs into prismatic loops (a crowdion could either be bound to an attractive solute, or pinned between two repulsive ones). This process takes place over timescales of seconds or more (depending on the dose rate), and is hence inaccessible to atomistic methods (MD).
In this work I apply the analytical Frenkel-Kontorova model [6,7] to crowdion defects lying in <111> strings containing a solute impurity. This was previously investigated in [8], where impurities having different masses and couplings were considered. These lead to quite different interaction potentials, and will be applied here to both transmutation impurities (same period, different group) and deliberately introduced alloying elements. The functional form of the crowdion-solute interaction potential is determined from the analytical model, and its parameters can then be fitted to the DFT results.
Once the interaction potentials are known, a coarse-grained simulation methodology such as Langevin dynamics [9] can be developed. This has the advantage of tracking only the “interesting” degrees of freedom (the defects themselves), accelerating the simulations by ~10⁷ cf MD and furthermore, does not require any rates to be calculated a priori cf kinetic Monte Carlo. Preliminary results will be presented and compared with transmission electron microscopy of radiation damage evolution in W and W alloys.
[1] Nguyen-Manh, D et al, Phys Rev B 73 020101 (2006)
[2] Dudarev, SL, et al, J. Nucl Mat 386 1 (2009)
[3] Olsson, P, et al, Phys Rev B 75 014110 (2007)
[4] Muzyk, M, et al, Phys Rev B 84 104115 (2011)
[5] Kong, X-S, et al, Acta Mat 66 172-183 (2014)
[6] Braun, OM and Kivshar, YS, “The Frenkel-Kontorova Model: Concepts, Methods, and Applications”, Springer (2004)
[7] Fitzgerald, SP et al, Phys Rev Lett 101 115504 (2008)
[8] Braun, OM et al, Phys Rev B 43 1060 (1991)
[9] Swinburne, TD, et al, Phys Rev B 87 064108 (2013)

Determining the swelling behavior of ferritic-martensitic alloys is important for predicting the safety and structural integrity of fast spectrum reactors. Self-ion irradiation experiments have been performed on ferritic-martensitic alloys HT9 and T91 to determine the microstructure evolution at 440°C to doses of 140 dpa and above using two different beam conditions and several He implant states. Irradiati