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. Irradiations were performed on both He implanted and unimplanted samples using either raster scanning or a defocused beam condition on a 1.7 MV Tandetron accelerator at the Michigan Ion Beam Laboratory. The effects of beam condition on bulk swelling were determined by examining the void distribution using transmission electron microscopy (TEM) in STEM mode. The defocused beam was found to enhance void nucleation and growth, regardless of helium implant condition. Additionally, dislocation microstructure and precipitation behavior were analyzed using TEM and a similar effect was found, that defocused beam enhanced the diameter of the dislocation loops as well as the Ni/Si precipitate diameter, but had less of an effect on the number density of the analyzed microstructural features. Finally, the results were interpreted within the context of a Fully Dynamic Rate Theory (FDRT) model, which determined that annealing of point defects between pulses of the rastered beam suppressed microstructural evolution.

Ni-Fe-Cr based fcc alloys are frequently used as critical structural materials in nuclear energy applications. However, despite extensive studies, fundamental questions remain regarding point defect migration and solute segregation to grain boundaries after high temperature irradiation. Specifically, a systematic study of the role of grain size and grain boundary character must be understood to develop insights into new material or processing routes that have enhanced radiation tolerance. In this study, a coupled experimental and modeling approach is used to examine the response of grain boundary character and grain size in model Ni-5Cr and Ni-18Cr alloys. After severe thermomechanical processing to induce a range of grain boundary types and grain sizes, heavy ion irradiation was carried out at 500°C. Post-irradiation analysis of specific grain boundaries using scanning transmission electron microscopy (STEM) in conjunction with x-ray energy dispersive spectroscopy to examine the interaction of irradiation induced defects, defect denuded zones, and Cr solute segregation as function of grain boundary character and grain size. The results highlight a variation in radiation induced segregation with grain boundary misorientation and plane normal with large variations for the coherent and incoherent twin grain boundaries. The experimental results were compared to kinetic rate theory modeling with gain boundary sink efficiency boundary conditions. The model and experimental radiation induced segregation profiles for the model Ni-Cr show good agreement as function of misorientation and when accounting for variations in grain boundary energy for the twin grain boundary plane or low angle grain boundaries. The study highlights a combined experimental and modeling approach to understanding the role of grain boundary character and grain size on critical nuclear energy Ni-Cr based alloys.
The authors gratefully acknowledge funding from the NSF, Division of Materials Research, MMN program, Grant No. 1105681, user project support by ORNL&’s CNMS, which is sponsored by the Scientific User Facilities Division, Office of BES, U.S. DOE, and Sandia National Laboratories, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. DOE NNSA under contract DE-AC04-94AL85000.

Nanostructured ferritic alloys are promising materials candidates for the next generation of nuclear reactors due to their ability to withstand high temperatures, high pressures, high neutron flux and especially, the presence of high concentrations of transmutation product helium. As helium diffuses through the matrix, large number densities of complex oxide nanoclusters, namely Y₂Ti₂O₇, Y₂O₃ and Y₂TiO₅, act as trapping sites for individual helium atoms and helium clusters. Consequently, there is a significant decrease in the amount of helium that reaches grain boundaries, mitigating the threat of pressurized bubble formation and embrittlement. In order to understand the helium trapping mechanisms of the oxides at a fundamental level, the interface between the nanoclusters and the iron matrix must be modeled. We present results obtained using density functional theory on the structural and thermodynamic properties of the Y₂Ti₂O₇-Fe interface in the presence of helium. In addition, helium bubbles of varying sizes have been introduced in order to observe the effects of a growing helium bubble.

Defect-driven diffusion of impurities is the major phenomenon leading to formation of embrittling nanoscopic precipitates in irradiated reactor pressure vessel steels. Here we will discuss solute-defect interactions and diffusion characteristics in dilute bcc Fe alloys. Flux coupling phenomena such as solute drag by vacancies and radiation induced segregation at defect sinks are systematically studied for a number of binary Fe alloys. The interaction and diffusion models are based on first principles calculations in the framework of density functional theory. Both kinetic Monte-Carlo simulations and the self-consistent mean field theory are applied in order to determine the tranposrt coefficients and the radiation induced segregation trends. We classify the different solutes depending on their character of interaction with the point defects. The consequences for radiation driven diffusion of these solutes are discussed.

Reduced-activation ferritic/martensitic steels are candidate materials for the first wall of a fusion reactor, because of the excellent resistance to radiation damage. It has been reported that void swelling in ferritic steels is promoted by the synergistic effect of transmuted hydrogen and helium, however, the effect of hydrogen and helium on the formation and the growth of vacancy-type cluster are not clear. In addition, dislocation motion in BCC iron would be affected by vacancy type clusters, which resulted in irradiation hardening.
In this study, in order to investigate the effect of hydrogen on the formation of vacancy cluster and the effect of vacancy cluster and vacancy-hydrogen complexes on dislocation motion, electron and ion irradiation experiments and the molecular dynamics (MD) simulation were performed for alpha-iron.
From the results of irradiation experiments, it is suggested that hydrogen would enhance the growth of cavity in the early phase of irradiation and the helium could assist the nucleation of cavity and hydrogen does the growth, which leading to a large swelling. The MD simulation indicated that edge dislocation and vacancy cluster seemed to have the attractive force and edge dislocation was bowed out when cutting through the cluster. On the other hand, screw dislocation and vacancy cluster also had the attractive force, however, screw dislocation showed cross slip when cutting through the cluster. Furthermore, vacancy-hydrogen complexes showed greater attractive force to edge dislocation compared to that of vacancy clusters, especially with hydrogen pair inside.

The crystalline defects produced under irradiation aggregate in clusters which play important role in the microstructural evolution of materials. Large time-space scale simulations for radiation embrittlement of fusion materials, based on radiation-induced defect and dislocation microstructure, generated by ions and/or neutrons, require valuable inputs for the growth of point-defect clusters. The dislocation loops at nanometric size are too small to be characterized by the experiment or too big to be investigated by a reliable energetic model as ab-initio. Empirical potentials give a good basis for self-interstitial clusters but the reliability of these potentials is continuously revised with the last advances in the field of ab-initio electronic structure calculations [1-4]. In this presentation we propose the development of an energetic model in Fe and W which is able to predict the relative stability of large self-interstitials clusters up to nanometric-size directly from ab initio calculations performed on small clusters. We will give particular attention to the relative stability of the traditional dislocation loops with <100> and ½<111> orientations in W as well as the C15 clusters, recently predicted by the DFT in Fe [5]. The theoretical findings will be compared with recent experiments.
[1] C. C. Fu, F. Willaime, and P. Odrejon, Phys. Rev. Lett., 92,175503, (2004).
[2] D. A. Terentyev, T. P. C. Klaver, P. Olsson, M.-C. Marinica, F. Willaime, C. Domain, and L. Malerba, Phys. Rev. Lett., 100, 145503, (2008).
[3] L. Malerba et al Journal of Nuclear Materials, 406, 19, (2010).
[4] M.-C. Marinica, L. Ventelon, M. R. Gilbert, L. Proville, S. L. Dudarev, J. Marian, G. Bencteux, and F. Willaime, Journal of Physics: Condensed Matter, 25, 395502, (2013).
[5] M.-C. Marinica, F. Willaime, and J.-P. Crocombette, Phys. Rev. Lett., 108, n025501, (2012).

The kinetic Monte Carlo (KMC) method is a useful tool for investigating radiation damage accumulation in materials such as nuclear reactor pressure vessel steels, and object KMC has often been used to simulate the time evolution of reactions involving point defects. Object KMC requires specifying values for essentially all the relevant physical parameters associated with these reactions, such as migration energies, binding energies, pre-factors for diffusion coefficients and reaction radii for interacting defects. These parameters are usually obtained from ab initio or molecular dynamics simulations, or estimated based on experimental data. Although vacancy-interstitial (V-I) recombination is a significant reaction since it removes both species from the system, there is not a fundamental basis for the value of the V-I recombination radius that has been used in the past. Previous computational studies have typically used values between 1.0 and 3.5 lattice parameters. We have carried out a systematic investigation of V-I recombination in bcc iron using the Self-Evolving Atomistic KMC (SEAKMC), a so-called on-the-fly method that has been shown to accurately simulate the kinetics of point defect reactions. Like molecular dynamics, it is essentially parameter-free with the system described by an interatomic potential. In this study the potential by Ackland and co-workers from 2004 was used. The simulation cell size was varied between 2000 and 8192 atoms to change the effective defect concentration and the temperature range of 473K to 773K was investigated. In each simulation, a single <110> dumbbell interstitial and a vacancy were separated by 5 lattice parameters and SEAKMC was used to determine their atomic trajectories until recombination occurred. Multiple runs were used to assess the statistical variation at different temperatures and defect concentrations (cell sizes). The V-I separation distance prior to the final atomic jump which resulted in recombination was taken as the recombination radius. Results indicate that the mean recombination radius in bcc iron is ~2.2 lattice parameters with a standard error of ~0.1 lattice parameters. Although the simulation time and mean square displacements depended on temperature and defect concentration, the recombination radius was essentially independent of these parameters for the conditions studied.

The mechanical integrity of many materials is impacted by the presence of hydrogen and even minute amounts of a few ppm may give rise to fracture and embrittlement. This is particularly true for metallic alloys designed to withstand high mechanical loads. From a modeling point of view, the description of the impact of H on mechanical properties is challenging since it is well known to interact with almost all microstructural and point defects. Thus, a quantitative description of how H impacts and interacts with the various defects requires to model large length- and timescales while also taking into account atomistic effects. In the talk it will be shown how ab initio calculations can be used to (i) identify the relevant mechanisms at atomic scale and (ii) to construct atomistically informed mesoscale/macroscale concepts that allow to model H embrittlement on the length and time scales relevant for experiment. It will be shown that even in non-hydride forming metals defect induced strain fields may induce and stabilize nano-hydrides. These nano-hydrides severely impact the structure of the defect, as well as its interaction with other defects and its kinetics and are shown to resolve many of the puzzling experimental findings.

Missing pellet surface (MPS) defects are local geometric defects that periodically occur in nuclear fuel pellets, usually as a result of the mishandling during the manufacturing process. The presences of these defects can lead to clad stress concentrations that are substantial enough to cause a through wall failure for certain conditions of power level, burnup, and power increase. Consequently, the impact of potential MPS defects has limited the rate of power increase or ramp rates in both PWR and BWR systems. Improved 3D MPS models that consider the effect of the MPS geometry can provide better understanding of the margins against PCMI clad failure. The Peregrine fuel performance code has been developed as a part the Consortium of Advanced Simulations of Light Water Reactors (CASL) to consider the inherently multi-physics and multi-dimensional mechanisms that control fuel behavior, including cladding failure by the presence of MPS defects. This paper presents an evaluation of the cladding stress concentrations as a function of MPS defect geometry. The results are the first step in a probabilistic approach to assess cladding failure during power maneuvers. This analysis provides insight into how varying pellet defect geometries affect the distribution of the cladding stress and fuel and cladding temperature and will be used to develop stress concentration factors for 2D and 3D models.

The UO₂ fuel grain size has a large impact on fuel performance, effecting the swelling, fission gas release, and creep. However, existing empirical models for grain growth do not consider the impact of initial porosity and fission gas bubble formation on the GB migration. As part of the ongoing effort funded by the US Nuclear Energy Advanced Modeling and Simulation Program to transition to fuel materials models based on microstructure rather than burnup, we determine the individual terms of an analytical model for grain growth using modeling and simulation at the atomistic and mesoscales. Molecular dynamics simulations are used to determine the GB mobility and energy. Mesoscale simulations using INL&’s MARMOT code determine the impact of the temperature gradient and the pinning force due to fission gas bubbles on the GB. This model is implemented in INL&’s BISON fuel performance, coupling to the fission gas release and densification models.

Grain boundary selective oxidation is one of the primary causes of intergranular corrosion and cracking for metallic alloys in light-water reactors. We combine quantum Density Functional Theory (DFT) and mesoscopic Poisson-Nernst-Planck / classical DFT to establish a mesoscale metal alloy oxidation model for grain boundaries at experimentally relevant length scales, focusing on Ni-Cr and Ni-Al binary alloys. The oxidation reaction involves vacancy-mediated transport of the alloying element and Ni to the oxidation front and the formation of stable metal oxides. DFT simulations revealed a strong dependence for the diffusion barriers of alloying element and Ni diffusion on the structure of the grain boundary and the nature of the minor element. These data were used in mesoscopic PNP/cDFT simulations to compare the details of the intergranular oxidation in Ni-xCr (x = 5, 10, 20, 30) and Ni-4Al alloys. Simulations revealed that oxidation of Ni-5Cr alloy leads to the formation of porous Cr₂O₃ oxide with channel-like pores along the grain boundary in agreement with experimental observations performed in high-temperature hydrogenated water. In addition, Ni enriches the grain boundary at oxide interface and is not found in the oxide region. The porosity of the oxide decreases with increasing Cr content, hindering access of oxidizing species to the oxide/metal interface for Cr levels >20%. The oxidation process is known to be associated with accumulation of vacancies and depletion of minor elements in the alloy ahead of the oxidation front. In Ni-5Cr, Ni vacancies are mainly concentrated at the alloy/oxide interface, while the concentration of Cr vacancies is significant several nanometers ahead of the oxidation front. In contrast to Ni-5Cr alloy, for Ni-4Al both Ni and Al are found in the oxide phase. The porosity of the oxide, spinel NiAl₂O₄, is also different. There are larger pores in NiAl₂O₄ along and smaller pores perpendicular to the grain boundary. The smaller pores with the average diameters of 0.6 nm are sufficiently large to allow transport of the oxidizing species to either side of the grain boundary. This may lead to the preferential oxidation of one side of the grain boundary and may explain an observed asymmetry in measured Al concentration across the grain boundary during corrosion in high-temperature hydrogenated water.

The evolution of the surface morphology and the near-surface structure of plasma facing components (PFCs) in nuclear fusion reactors is impacted significantly by the implantation of helium (He) atoms. Tungsten (W) is an important PFC material due to its thermomechanical properties. In tungsten, such interstitial He atoms are very mobile and aggregate to form clusters of different sizes. The smallest of these helium clusters, with up to seven helium atoms, also are mobile and their diffusional transport mediates the evolution of surface morphology and the sub-surface gas bubble structure and dynamics.
In this presentation, we report the results of a systematic atomic-scale analysis of the interactions of small mobile helium clusters with sinks and the cluster reactions near these sinks in tungsten. Sinks that have been investigated include free surfaces, grain boundaries (GBs), and junctions where GBs intersect free surfaces. Molecular-statics (MS) simulations based on reliable many-body interatomic potentials for relaxation of helium clusters near sinks are used to obtain the potential energy profiles of the clusters as a function of their center-of-mass distance from a sink. Elastic interaction potentials based on elastic inclusion theory provide an excellent description of the computed cluster-sink interactions. The key parameter in the elastic models is the sink segregation strength, which increases with increasing cluster size. These cluster-sink interactions are responsible for the migration of small helium clusters by drift and for helium segregation on surfaces and GBs in tungsten. Such helium segregation on sinks is observed in large-scale MD simulations of helium aggregation in model polycrystalline tungsten at 933 K upon helium implantation. As the clusters migrate toward the sinks, trap mutation (TM) and cluster dissociation reactions are activated at rates higher than in the bulk; near surfaces, TM produces W adatoms and immobile complexes of helium clusters surrounding W vacancies. These reactions are identified and characterized based on analysis of many molecular-dynamics (MD) trajectories for each such mobile cluster near surfaces and GBs. For helium clusters near W surfaces, TM is the dominant reaction except for 4- and 5-member clusters near W(100) where cluster partial dissociation following TM dominates. We find that there exists a critical cluster size beyond which formation of multiple W adatoms and vacancies in TM reactions is observed. For the GBs examined, TM reactions were found to be dominant for larger than 4-member clusters, but no critical mobile cluster size has been found for the formation of multiple vacancies in TM reactions. The identified cluster reactions are responsible for important structural, morphological, and compositional features in plasma-exposed tungsten, including surface adatoms, near-surface immobile helium-vacancy complexes, and retained helium content.

We present a novel phase field (diffuse-interface) model of void growth in irradiated materials. Since the void surface is inherently sharp, diffuse interface models for void growth must be constructed in a way to make them consistent with the sharp-interface description of the problem. Therefore, we first present the sharp-interface description of the void growth problem and deduce the equation of motion for the void surface. With the sharp-interface analysis in mind, we then construct a phase field model of type C, which couples Cahn-Hilliard and Allen-Cahn equations, to properly represent the void growth phenomenon. Such model is able to capture the diffusion of point defects in the bulk by modified Cahn-Hilliard equations and account for the reaction of point defects at the void surface via an Allen-Cahn equation. We carry out a formal asymptotic matching between the diffuse-interface and sharp-interface models. The asymptotic matching demonstrates that the kinetic equations of the phase field model recover the same equation of motion for the void surface that was deduced from the sharp-interface model. Furthermore, the asymptotic matching enables us to establish a direct relationship between the model parameters in the diffuse-interface model and the physical properties of materials. Sample results for void growth in a single component metal based on sharp and diffuse interface models are presented. We found out that the reaction of point defects with the surface strongly affects the overall growth rate of voids in irradiated solids. This material is based upon work supported as part of the Center for Materials Science of Nuclear Fuel, an Energy Frontier Research Center funded by the U.S. Department of Energy, Office of Basic Energy Sciences under award number FWP 1356, through subcontract number 00122223 at Purdue University.

We present a 3D phase field model for investigating the grain growth process in porous ceramics. The grain growth process in ceramics is complicated by the interaction between the pores and the grain boundaries. As such, in addition to grain boundary migration, the model takes into consideration pore migration via surface diffusion and hence pore coalescence. Therefore, the model is able to fully capture the microstructure evolution in porous ceramics. All model parameters are obtained from material properties. Application of the model to uranium dioxide shows that the grain growth in this material is sensitive to the level of porosity. The kinetics of grain growth in uranium dioxide changes from boundary-controlled kinetics to pore-controlled kinetics. Furthermore, the model captures the pore breakaway phenomenon which is observed experimentally. The effects of porosity, temperature, mobilities, and initial microstructure on the grain growth process were investigated .The model results agree well with grain growth experiments. This research was performed as a part of the Energy Frontier Research Center, Center for Materials Science of Nuclear Fuel funded by the U.S. Department of Energy, Office of Basic Energy Sciences.

Stable nanoclasters have been observed in Fe-based alloys fabricated by mechanical alloying. These so called oxide-dispersion strengthened alloys (ODS) demonstrate improvement of creep resistance which makes them a strong candidate for usage in applications at high temperature and in extreme environments. We investigate the stability of ODS (precipitate) particles and the distribution of thermal defects (vacancies, antisites) using a combination of density functional theory (DFT) and the Grandcanonical formalism. In this approach we describe the free energy of the ODS alloy as a weighted sum of the Y₂O₃ and Fe compound free energies. The resulting atomic distribution is obtained from the minimization of Grandcanonical potential with an additional constraint for conservation of number of particles in the Y₂O₃ - Fe system. Such an approach neglects the Y₂O₃/Fe interphase surface effects and corresponds to the limit of large ODS particles. The energies of vacancies and anti-sites in perfect Y₂O₃ and Fe crystals cells containing 40 and 54 atoms respectively were calculated using DFT. Analysis of the calculated defect concentrations at 2000K in the compound containing 0.5% Y₂O₃ in an Fe-matrix shows that the most favorable defects are Fe vacancies, oxygen interstitials (octahedral site), and Fe-Y antisite defects. In atomic units, the vacancy concentration is 10^-4, oxygen interstitial concentration is ~10^-7, and approximately 10^-9 Fe and Y atoms have exchanged positions. The overall conclusion is that yttria is very stable in the Fe matrix. Next edition of the current model will take into account the Y₂O₃/Fe interface and yttria particle size distribution.
This research was sponsored by the Materials Sciences and Engineering Division, Office of Basic Energy Sciences, US Department of Energy, and 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, US Department of Energy.

The CALPHAD method is used to describe the thermodynamic properties and phase relations in the U-M-O system where M = Gd, La, and Th. A compound energy formalism (CEF) model for fluorite UO_2±x is extended to represent the complex U_1-yM_yO_2±x phase. The lattice stabilities for fictive GdO₂ and LaO₂ fluorite structure compounds are calculated from density functional theory (DFT) for use in the CEF for U_1-yM_yO_2±x while U⁶⁺ is introduced into the cation sublattice of that model to better reproduce phase relations in the U-La-O system at high La compositions. Tentative Gibbs functions and CEF representations for the fluorite derivative rhombohedral phases were developed and the two-sublattice liquid model (TSLM) was used to describe the melt. Equilibrium oxygen pressures over U_1-yTh_yO_2±x were obtained from thermogravimetric measurements and used together with those reported in the literature, phase relations, and other experimentally determined thermodynamic values to fit adjustable parameters of the CEF and TSLM. The models can be extended to include other actinides and fission products to develop higher order multi-component thermodynamic descriptions.

The formation of zirconium hydrides in zirconium alloys is an important aspect of the microstructural evolution of fuel-pin cladding materials in light water reactors. To maximize fuel performance and lifetime, a more in-depth and predictive understanding of hydrogen effects on the cladding is required than is currently modeled in fuel performance codes. A phase field modeling framework, Hyrax, has been developed to examine the precipitation and growth of zirconium hydrides within zirconium using the finite element framework MOOSE. Hyrax incorporates into a phase field model a nucleation algorithm based on classical nucleation theory, heat conduction, and hydride/zirconium misfit strain to examine precipitation and growth phenomena in single crystal zirconium. A CALPHAD-based free energy functional provides realistic energetics of the system and of hydride nucleation driving force. The temperature dependence, applied stresses, and hydrogen flux conditions have been incorporated into the CALPHAD free energy functional and other model components, enabling Hyrax to simulate a wide range of technically relevant operating conditions. The simulations reproduce the experimentally observed hysteresis of the terminal solid solubility of the hcp-Zr solid solution upon heating and cooling, and the results are consistent with the hypothesis that the hysteresis arises from elastic strain energy due to matrix/precipitate misfit. We also study the effect of applied stresses on the phase stability and document the sensitivity of the hydride nucleation rate on parameters including applied stress, temperature, supersaturation, and hydride/zirconium interfacial energy.

Beryllium is a leading candidate for plasma facing and neutron multiplier components in fusion reactor designs. Understanding the microstructural changes caused by common impurities such as Al and Fe, and their interaction with irradiation and fusion products is crucial for the development of high tritium release, radiation resistant Be-alloys.
The current work uses ab-initio simulations coupled with lattice dynamics calculations and the Bragg-Williams method for disordered phases, to investigate the following compounds: FeBe₂, FeBe₅, the ε phase (often referred to as FeBe₁₁), AlFeBe₄ and the Be(Al) and Be(Fe) solid solutions.
We provide a new crystallographic structure for the Be-rich ε compound, which was until now unknown. Off-stoichiometry was also investigated, and found to be important for the ε phase and for FeBe₅. We also predict that the FeBe₅ is thermodynamically unstable at temperatures below ~1250K, however small addition of Al stabilise the FeBe₅ phase over the other intermetallic compounds. Furthermore, FeBe₅ starts to exhibit disorder above 500K and becomes completely disordered above 1500K.
Larger amounts of Al lead to the formation of a ternary phase based on the same structure as FeBe₅. This phase was previously termed AlFeBe₄, however, the current study shows that it exhibits complete disorder on the Al/Fe sublattice, therefore we propose the name (Al,Fe)₂Be₄.
Off-stoichiometry was also investigated, and found to be important for the ε phase and for FeBe₅. Implications with respect to radiation tollerance and accommodation of fusion products and other impurities are discussed.

Materials utilized in nuclear systems are subjected to extreme environments of high temperature, corrosion and intensive radiations, and advancements in nanotechnologies may play a crucial role to develop advanced materials with extended performance that can sustain such extremes. In this talk, key materials challenges will be discussed with the emphasis of accident tolerance for both fuel and cladding materials, and effective nuclear waste managements. The potential opportunities in applying nanotechnologies for designing advanced nuclear materials will be highlighted based on key concepts: (1) nanostructured materials and engineered microstructures for enhanced thermal-mechanical properties and radiation tolerance; and 2) nanomaterials for effective nuclear waste managements. Issues limiting the application of nanostructured materials in nuclear environments are also discussed with the focus on the structural integrity and radiation tolerance of nanomaterials under coupled extremes of high temperature and intensive radiations.

Nanoporous metallic materials offer superior mechanical, catalytic and optical properties showing great potential for applications such as actuators, sensors and future energy devices. Recent studies on radiation damage of nanoporous materials suggest that the abundance of surfaces, which are perfect sinks for defects, and the combined effect of ligament size, irradiation temperature, and defect migration, are decisive factors in determining their radiation behavior. Our previous experimental results indicate that radiation damage in nano-scale gold foams (np-Au) manifests itself mainly by the formation of stacking fault tetrahedra (SFTs). In this work, we explore the effect of cascade damage on np-Au caused by irradiations with He, Ne and Kr ions at room temperature, and at high, intermediate, and low ion dose-rates, to a total dose of 1 dpa. We aim at determining the boundaries of our recently defined ‘window of radiation tolerance&’, which relates material features such as ligament size and diffusion properties to irradiation properties such as pka energy, dose rate, and temperature [E. Bringa, et al., Nano Letters 12, 3351 (2012)]. We conclude that in the region of the parameter space we explored, we effectively cross a boundary separating regions of endurance and damage.

Advanced fuels with enhanced safety margin and accident tolerance are of vital importance for safe operation of current light water reactor fleet and the development of future reactor and transmutation systems. The current UO₂ fuel for LWRs has low thermal conductivity which will be further reduced upon burn up. The intrinsically-low thermal transfer efficiency results in large temperature gradients within the fuel pellet and high center line temperature. Thermal swelling of the pellets and fission product accommodation and retention limit the lifetime of UO₂ fuel in the reactor, presenting the great challenges for extending burn-up, effective utilization of nuclear resources and minimization of nuclear waste accumulation. Fukushima event further posts a critical need in developing accident tolerant fuels that can tolerate loss of active cooling for a considerable time span after accident or improve fuel performance during regular operation. Current efforts on accident tolerant fuels mainly focus on improvement of thermal-mechanical properties and fission product retention capability. Different fuel forms are proposed including the incorporation of heterogeneous fillers (SiC, diamond, and carbon nano-tube) or doped fuels in controlling fuel microstructures.
In this work, we demonstrate graphene-based composites for accident tolerant fuels in which the highly thermal conductive and mechanically robust graphene nano-platelets can greatly improve the thermal conductivity of the UO₂ fuel matrix and mechanical properties. UO₂ /graphene composites with various loading of nano-platelets (1, 3, and 5 wt.%) were prepared by high energy ball milling (HEBM) and consolidated into dense fuel pellets by spark plasma sintering (SPS), a state-of-art fast assisted sintering technology (FAST). The microstructure and physical density of the SPS-densified composite pellets at different temperatures (e.g., 1500 and 1600 ^oC) and duration (5, 8, and 20 mins) were investigated, and the thermal physical properties including thermal conductivity were measured. The two-dimensional geometry of graphene nano-platelets under uniaxial hydrostatic pressure during SPS process leads to a unique lamellar microstructure. Thermal transporting properties (thermal diffusivity and conductivity) are found to be anisotropic in the pellet with a 1-2 fold enhancement in radial direction; whereas the out-of-plane thermal properties of the fuel pellets remains un-degraded. The extraordinary high in-plane diffusivity of sintered pellets can improve the heat transfer efficiency along the radial direction of the fuel pellets and thus greatly reduce fuel temperature gradient/center line temperature. The stability of the graphene nano-platelets under relevant operation conditions was analyzed, and the potential of graphene-based UO₂ composite as accident tolerant fuels was evaluated.

Tungsten laminate materials constructed from rolled sheet are promising candidates for structural applications in future fast neutron fusion and fission nuclear systems. However their response under irradiation has not been studied. High-energy neutron damage will cause both cascade damage and also composition change through transmutation, producing up to 10% Re after in reactor service. In this study, W-5%Re was used as an analogue to such effects, and was ion-implanted to assess effects of radiation damage.
Rolled tungsten 5 wt% rhenium sheet was studied in two microstructural variants: (a) as received with a high dislocation density (mean value of 1.4×10¹⁴lines/m²), measured using HR-EBSD, and pancake shaped grains with a thickness ofasymp;200nm and (b) annealed at 1400^oC for 24 hours to produce equiaxed grains with average grain size of asymp;90 µm and low dislocation density (with a mean value of 4.8×10¹³ lines/m²). Both materials were ion implanted with 2MeV W+ ions at 300^oC to damage levels of 0.07, 0.4, 1.2, 13 and 33 displacements per atom (dpa). Nanoindentation was used to measure the change in hardness after implantations. Irradiation induced hardening saturated in the as-received material at an increase of 0.4dpa from the unimplanted hardness of 8GPa at 0.4dpa. In the annealed material saturation does not occur by 13dpa and the hardness change of 1.3GPa from the unimplanted hardness of 6.2GPa was over four times higher. At 33dpa both material types showed a further increase in hardening of 2.1GPa (as-received) and 3.2GPa (annealed). Atom probe tomography showed no clustering of Re at 13dpa and below. At 33dpa clusters of asymp;4nm diameter with a rhenium concentration of asymp;11% were seen in both material types. In both cases the number density and volume fraction are similar at asymp;3100 x1000/µm³ and volume fraction of asymp;13%.
These differences in radiation response are likely to be due to the high damage sink density in the as-received microstructure in the form of dislocation networks, as even in the as-received material the average grain size is too large to provide sufficient sinks. Initially this provides a large sink network for radiation damage resulting in less hardening in the rolled material. However at 33dpa the formation of rhenium clusters occurs at similar levels in both material conditions. These dominate the hardening mechanisms and result in secondary hardening at high damage levels. This work demonstrates the advantage of using such nanostructured tungsten sheet in composite materials for structural applications as they will have improved radiation resistance as compared to bulk tungsten products, at low damage levels. At higher damage levels changes in chemistry may result in radiation induced segregation dominating and extreme hardening and the possibility of embrittlement occurring. This study also highlights the danger of using idealized annealed microstructures for radiation damage studies due overestimations of hardening responses.

Nanocrystalline oxides are of high interests for a wide range of applications due to their exceptional size-dependent materials properties. The effective use of nanostructured oxides in nuclear related applications requires better understanding of these materials performance in harsh environments, including high temperature and extreme radiation. In particular, nanostructured oxides are considered as potential candidates to meet the demand for advanced fuels and cladding materials that can withstand extreme radiation environments with improved accident tolerance over a long period of time, and with improved performance in advanced nuclear energy systems. Cubic ceria (CeO₂) and zirconia (ZrO₂) are well known ionic conductors that are also isostructural with urania, plutonia, and thoria-based nuclear fuels. Understanding irradiation response of ionic-covalent materials is important for advanced nuclear energy systems.
Ion beam processing is an effective approach to tailor size-dependent material properties of oxide-based nanomaterials. Grain growth of nanocrystalline materials is generally thermally activated, but can also be driven by irradiation at much lower temperature. Under ion irradiation, defect production and ionization effect lead to effective modification of interface volume in nanocrystalline ceria and zirconia. Experimental results have shown that both high electronic energy loss and nuclear energy loss lead to disorder and radiation-induced growth of the crystallite size is a function of total energy deposited. Atomistic simulations by adding high levels of disorder in the simulation cell have revealed fast grain boundary (GB) movements due to the present of high-level disorder in the close proximity to GBs, and the results is in good agreement with our the experimental results. The coupling of energy deposition to the electronic and lattice structures should both be taken into consideration when engineering nanostructural materials.
This work was supported by the Materials Science of Actinides, an Energy Research Frontier Center supported by the U.S. Department of Energy, Basic Energy Sciences.

Grain boundaries have been shown to be effective sinks for radiation-induced point defects and associated dislocation loops. This observation suggests that nanocrystalline materials are good candidates to resisting mechanical property degradation in irradiation environments. However, nanometer-scale grains tend to be highly unstable with respect to thermal annealing. Recently, several alloy materials have proven to be resistant to grain growth at temperatures up to 400 0C. Are these materials also resistant to irradiation induced creep? As long as temperature is held below that at which diffusional creep initiates, this strategy may prove effective if radiation-enhanced diffusion enables point defect to reach the grain boundaries. In this preliminary study, we explore electroplated NiW films of 100 nm thickness with grain sizes as small as 3 nm. Both cathodic and pulsed plating methods on electropolished Cu substrates are used to incorporate W into the Ni films and control grain size. An anneal above 125 0C tends to relax the grain boundaries, however, resulting in higher grain boundary ordering. Therefore, we self-irradiate films with Ni ions that have and have not been annealed to determine their relative effectiveness in gettering defects. Analysis is conducted by high resolution transmission electron microscopy.

Oxide dispersion strengthened ferritic alloys have superior radiation tolerance and thus become appealing candidates as fuel cladding materials for next generation nuclear reactors. In this study we constructed a model system, Fe/Y₂O₃ nanolayers with individual layer thicknesses of 10 and 50 nm, in order to understand their radiation response and corresponding damage mitigation mechanisms. These nanolayers were subjected to in situ Kr ion irradiation at room temperature up to ~ 8 displacements-per-atom. As-deposited Y₂O₃ layers had primarily amorphous structure. Radiation induced prominent nanocrystallization and grain growth in 50 nm thick Y₂O₃ layers. Conversely, little crystallization occurred in 10 nm thick Y₂O₃ layers implying size dependent enhancement of radiation tolerance. In situ video also captured grain growth in both Fe and Y₂O₃ and outstanding morphological stability of layer interfaces against Kr ion irradiation. This research is supported by NSF-DMR-Metallic Materials and Nanostructures program.

Abstract
Grain boundaries and interfaces are known to promote radiation tolerance in materials in a nuclear reactor. We offer a novel strategy that can trap radiation-induced point defects between grain boundaries and inhibit them from evolving into extended defects. With optimized superlattice structures, we show using molecular dynamics (MD) simulations that spatial control of defects can be enforced in β-SiC in a radiation environment. We have conducted a series of non-equilibrium radiation cascade simulations both on single crystal SiC and superlattices comprising of several grains boundaries using a modified Tersoff interatomic potential that has been benchmarked to a host of thermo-mechanical and defect properties. While the layered superlattice structures do not reduce the number of defects appreciably, the spatial spread of the defects is significantly altered. In a single crystal, the defects are predominantly distributed in the longitudinal direction - in the direction of the knock - while they are distributed in the transverse direction - perpendicular to the direction of the knock - for the layered structures. Thus we show that layered superlattice structures can be advantageously employed to control the spatial defect distribution in a radiation environment.

Swift heavy ions (SHIs) induce a cylindrical region of structural transformation known as ion track. Ion tracks in SiO₂ can be used to change its refractive index or as a mean to induce anisotropy for etching [1]. Furthermore, it was recently found out that ion irradiation can be used to induce a shape transformation in metal nanocrystals (NCs) that are embedded in silica. Spherical NCs elongate along the ion beam direction and are shaped into nanorods or prolate spheroids. The phenomeon can be exploited to produce large arrays of equally aligned nanocrystals, which is difficult to achieve otherwise. The mechanism by which this transformation occurs is unclear.
The effects of SHI impacts in materials are poorly understood on the atomic scale and a predictive theory is essential for the controllability of the SHI-induced structural changes. We model these using the so-called Two-Temperature Molecular Dynamics method [2-3]. The simulations help us to explain and provide an insight into the fundamentals of ion-solid interactions.
[1] B. Afra et al., J. Phys.: Condens. Matter 25, 045006 (2013)
[2] A.A. Leino, O.H Pakarinen, F. Djurabekova, K. Nordlund, P. Kluth and M. C. Rigway, Materials Research Letters 2, 37 (2014)
[3] A. A. Leino, S. Daraszewicz, O H. Pakarinen, F. Djurabekova, K. Nordlund, B. Afra and P. Kluth, Nucl. Instrum. Methods Phys. Res. B 326 289 (2014)

Corrosion of cladding progresses during a light water reactor operation, which causes deterioration of the fuel performance. It is reported that second phase particles (SPPs) around metal / oxide layer interface influence the corrosion mechanism. It remains unknown about effects of the SPPs on the corrosion resistance. In this study, mechanical properties of Zr(Fe,Cr)₂, Zr₂(Fe,Ni), and β-Nb(-Zr), which are SPPs of Zircaloy-2, Zircaloy-4 and Zr-Nb alloys, were evaluated. Additionally, residual stress in oxide layer on the corroded zirconium alloys was measured for estimation of the corrosion resistance.
The samples of SPPs were prepared by arc melting and heat treatment. The crystal structure and composition were analyzed by XRD and SEM/EDX. Young's modulus was measured by ultrasonic pulse echo method. The Vickers hardness was measured by Vickers hardness test. As samples for corrosion test, Zr-Nb, Zr-Fe-Cr, and Zr-Fe-Ni alloys were prepared by arc melting. Residual stress in the oxide layer was calculated by XRD analysis after corrosion test in an autoclave.
Young's modulus of β-Nb(-Zr) alloy, Zr(Fe,Cr)₂, and Zr₂(Fe,Ni) slightly changed in proportion to the composition. The Vickers hardness values of Zr(Fe,Cr)₂ and Zr₂(Fe,Ni) were significantly higher than that of a-Zr. Residual compressive stress in oxide layer on Zr-Nb alloy was highest in the alloys, which indicated that the Zr-Nb alloy had a superior corrosion resistance.

The direct laser metal deposition (DLMD) technology with co-axial powder injection was used to fabricate a complex titanium aluminide structure of Ti based alloy reinforced by Al₂O₃ ceramic of micron and nano size particles. The aim of the study was to demonstrate the possibility of producing Ti_xAl_y intermetallic phases in remelting powder mixtures with strain-hardening ceramic inclusions in the course of the single-step DMD process. Besides, relationships between the main laser cladding parameters and the intermetallic phase structures of the built-up objects were studied. In our research we applied optical microscopy, X-ray analysis, microhardness measurement and SEM with EDX analysis of the laser-fabricated intermetallics.

Various properties of structural and functional alloys including alloys for the nuclear technology are changed by the precipitations of solute atoms. We produced Fe precipitates in supersaturated Cu-1at.%Fe and Cu-3at.%Fe alloys by means of energetic particle (electrons or ions) irradiation or thermal aging and observed their local structure around Fe atoms by using the extended x-ray absorption fine structure (EXAFS) measurements at a synchrotron radiation facility. To investigate the effects of Fe precipitates on hardness and magnetic properties, we performed the micro Vickers hardness and SQUID#12288;measurements, respectively. From the Vickers hardness measurements for the thermally aged specimens at 873 K and 973 K, age-hardening curves having the maximum value for 5 hour-aging and 10 minute-aging were confirmed, respectively. This result implies that, Fe precipitates were produced by thermal aging. Next, we confirmed from EXAFS measurements that Fe precipitates for the specimens aged at 873 K had the fcc structure irrespective of aging time. On the other hands, the EXAFS FT-spectra for the specimens aged at 973 K for several days showed that Fe precipitates had the bcc structure. Furthermore, these specimens aged at 973 K for several days had a large value of the saturated magnetization. The present experimental results suggest that at the initial stage for the growth process of Fe precipitates, they exist as γ-Fe that has fcc structure. However, if they grow enough, they will transform from γ-Fe to α-Fe.
Moreover, we will also present the effects of deformation on the structure for Fe precipitates. The irradiation effects will also be discussed in the meeting.

Zirconium hydride (ZrHx) is currently expected as a neutron shield material for fast reactors. Therefore, the thermal and mechanical properties of ZrHx have been widely studied. However, the properties of zirconium deuteride (ZrDx) have been scarcely studied. On the other hand, since chemical properties of Zr and hafnium (Hf) are quite similar, Zr contains a few percent Hf generally. Therefore, it is very important to evaluate the effect of Hf content on the properties of ZrDx.
In the present study, fine bulk samples of d-phase ZrDx with various Hf contents were prepared and their thermal and mechanical properties were investigated. We examined the phase states and the microstructure of samples by means of XRD and SEM/EDX analyses. In the temperature range from room temperature to 673 K, the heat capacity and the thermal diffusivity were measured and the thermal conductivity was evaluated. The Vickers hardness and sound velocities were measured at room temperature, and the elastic modulus was calculated from the measured sound velocities. Effects of Hf content on the properties of ZrDx will be discussed.

Abstract:
Thorium has long been considered a possible source of fuel for use in power generating nuclear reactors. However, likely due to the already widespread use and familiarity of uranium oxide (UO₂) as a fuel, most of the focus on thorium centered on using thorium oxide (ThO₂) in reactor configurations comparable to those already in operation. As a result, there has not been much attention placed on the possibility of using thorium metal as the principal fertile material in reactor fuel. The prospect of using a thorium metal based fuel may provide some possible distinct advantages over the use of a thorium oxide fuel. Such as, a smaller necessary critical mass and potential changes in neutron moderation, which could potentially allow for a smaller and more economically feasible reactor design.[1]
The goal of this project is to develop accurate and precise computational models of the microstructural and mechanical properties of thorium metal for use in comparison of the two possible fuel types (Th-metal & ThO₂). The Vienna Ab Initio Simulation Package (VASP) was used to construct the α-thorium unit cell (fcc lattice structure) and perform various simulations of its elastic and mechanical properties under the Density Functional Theory framework (DFT).
Conjugate Gradient and RMM-DIIS methods were first used to relax the unit cell and lattice structure. Then the stress-strain relation and elastic behavior of the cell were determined by providing small displacements to each atom. Lastly, using these elastic constants in conjunction with the Voigt-Reuss-Hill approximations, the metal&’s Bulk modulus, Shear Modulus, Young&’s Modulus, and Poisson&’s ratio were calculated. Every step was performed using a pre-constructed LDA thorium pseudopotential and then again using a PBE-GGA pseudopotential, both supplied within the VASP software.
Comparing the simulation results with experimental literature values showed that both pseudopotentials slightly under predicted the volume and lattice constants of the unit cell, and produced mixed over/under predictions of the elastic behavior. Overall, however, both pseudopotentials agreed relatively well with experimental values, and the PBE-GGA slightly outperformed the LDA.[2,3]
References:
[1] M. Lung, “A Present Review of the Thorium Nuclear Fuel Cycle”, Nuclear Science and Technology, European Commission (1997)
[2] P. E. Armstrong, O.N. Carlson, J.E. Smith, “Elastic Constants of Thorium Single Crystals in the range 77-400K”, Journal of Applied Physics 30, 36 (1959)
[3] J. H. Frye, “Interim Report on Metallurgy of Thorium and Thorium Alloys”, Oak Ridge National Laboratory (1951)

From a nuclear forensic perspective, relating the microstructure of interdicted uranium alloys to the processing steps employed to create such alloys may help in establishing provenance of such interdicted materials. When U-Nb cools to room temperature, it undergoes a two phase transformation. First, the gamma phase (body-centered cubic) transforms to gamma knot (tetragonal). Second, the gamma knot transforms to alpha double prime (monoclinic). Such a transformation is similar to the diffusionless beta to omega phase transformation in Ti- and Zr-rich alloys. We examine manufacturing techniques in surrogate materials. In our procedure, we first arc melted various compositions of Zr-Nb and Ti-Nb. In both cases, niobium had various atomic percentage amounts all less than 20%. We used the method of gas tungsten arc welding (GTAW) to melt the samples. We then annealed the samples for 2 hours. Next, we hot and cold rolled the samples to approximately 4mm and 2mm thick, respectively. The arc melting, annealing, and rolling procedures were performed in order to match standard industrial processing techniques. We then performed various characterization techniques on the samples including, x-ray diffraction (XRD), energy dispersive x-ray spectroscopy (EDS), and electron backscatter diffraction (EBSD). We gathered data and analyzed the crystal structure, compositions, phases and textures of the samples. Such experiments serve as input to inverse process modeling and can provide useful information about materials provenance during nuclear forensics investigations.

Metallic uranium and uranium alloy properties are important to understand for use in the new generation of advanced fast reactors as well as for defense-related nuclear fuels. Uranium has a complex electronic structure and low symmetry crystal structures that lead to complex deformation behaviors. These deformation behaviors change the material&’s microstructure and ultimately its physical properties as well as provide insight into the physical processes that the uranium metal has undergone for application in nuclear forensics.
Uranium is a transition metal that has three distinct stable solid phases (α,β and γ). Uranium&’s ground state phase is the α (face-centered orthorhombic) phase. As the temperature increases, uranium will transition to the β (tetragonal) and the γ (body centered cubic) phases respectively. The α and β phases of uranium have unwanted an-isotropy of expansion, but the γ phase has seemingly isotropic expansion which is desirable in reactor environments. Uranium&’s transition temperatures from α to β, β to γ, and γ to liquid are 940.85K, 1047.95K, and 1405.95K respectively.
There have been a limited number of studies on the deformation mechanisms for uranium. In previous uranium deformation research, the common slip and twinning systems for α-uranium were well documented. However, there were far fewer publications that reported the critical resolved shear stress of these deformation mechanisms. The β and γ phases of uranium have been studied even less.
Recently there have been a few continuous models that attempted to replicate the deformations seen in α-uranium, but the fundamental deformation tendencies and yield stresses were often approximated or assumed. Therefore, in an attempt to fill the lack of experimental knowledge and allow for detailed analysis of uranium deformation behavior, a nanoscale plasticity molecular dynamics simulation is used to simulate and understand the plastic deformation behavior and kinetics.
Deformation in α-uranium can be caused by either mechanical stresses or heating and cooling, due to the large anisotropy of thermal expansion. The ground state α-uranium phase is a low symmetry phase that has a high propensity for twinning, with over 40 twinning modes and very few slip systems. The twin lamellae found in α-uranium grains tend to be numerous and very thin, rarely larger than 5 mu;m. A nanoscale plasticity simulation allows for a variety of slip and twinning system of α-uranium to be analyzed and allows for the study of how each contributes to uranium&’s deformation behaviors.

We present preliminary results regarding physical property measurements and aqueous corrosion tests that we performed on titanium aluminum carbide coatings deposited onto zirconium (Zr) alloy substrates. We are working to develop corrosion resistant coatings as a means to modify zirconium alloy nuclear cladding. The purpose of the coating is to slow the rate of cladding oxidation during an off-normal temperature excursion in a light water reactor. We report here on the following coating architectures: (1) Ti₂AlC (1.1 µm) / Zr; (2) Ti₂AlC (1 µm) / Ti (100nm) / Zr; and (3) Ti₂AlC (1 µm) / TiC (100nm) / Zr. These coatings were deposited at ambient temperature using magnetron sputtering. The crystal structures of the as-deposited carbide coatings were determined using grazing incidence X-ray diffraction (GIXRD) on plan-view Zr-alloy coupons. We found that the crystal structure of our coatings is consistent with titanium carbide (TiC) with some Al dissolved into the TiC, rather than the MAX phase compound, Ti₂AlC, that we intended to make. We also performed cross-sectional scanning electron microscopy (SEM) measurements on our coated Zr-alloy samples, in order to assess the microstructures and chemistries of our coatings. Finally, we performed corrosion tests in water on each of the samples described above. These tests were conducted in an autoclave at 360°C for 3 days. We will report on the results of these tests in this presentation.

Silicon carbide is a material proposed for structural components in fusion reactors and cladding materials for gas-cooled fission reactors due to its high temperature stability, chemical inertness and small neutron capture cross-section. When used in nuclear applications, this material can be exposed to various kinds of radiation, which cause the formation of numerous defects, such as vacancies and vacancy clusters, which induce a significant evolution of the material physical properties. Moreover, the vacancies can trap insoluble fission products, in particular fission gases, which can further affect the properties of the material.
One of the experimental techniques permitting the investigation of vacancy-type defects is positron annihilation lifetime spectroscopy (PALS), which consists in recording the radiation emitted at the beginning and the end of life of positrons in a material and deducing the positron lifetime and the properties of the electrons with which they have annihilated. Vacancies can trap positrons and can therefore be detected by monitoring e.g. changes in the lifetime of positrons in the material. To identify the types of defects present in the examined materials, however, comparison with calculated positron lifetimes or with results of other characterization techniques is required.
In this work the temperature dependence of 12 MeV proton irradiation induced point defects in 6H-SiC was studied by means of isochronal annealing followed by both positron annihilation lifetime spectroscopy and electron paramagnetic resonance (EPR) measurements. To help in the interpretation of the experimental results, the formation energies and positron lifetimes of various vacancy clusters were calculated [1,2,3]. The combination of the experiments and calculations enabled the identification of a negative silicon vacancy in the as-irradiated samples, with the lifetime of 218 ps, which is annealed between 400^#9702;C and 700^#9702;C. This process involves vacancy migration and formation of the V_C+V_Si cluster, with a lifetime of 235 ps.
[1] J. Wiktor, G. Jomard, M. Torrent and M. Bertolus, Phys. Rev. B 87, 235207 (2013).
[2] J. Wiktor, G. Jomard, M. Bertolus, Nucl. Instrum. Meth. B 327, 63 (2014).
[3] J. Wiktor, X. Kerbiriou, G. Jomard et al., Phys. Rev. B 89, 155203 (2014).

Zirconium-based alloys have long served as fuel rod cladding due to their good corrosion resistance and low neutron cross section. Nevertheless, corrosion of Zirconium-based alloys has been a major obstacle, as it is a limiting factor in both increasing reactor burnup and in the development of generation IV reactors. Although these materials have been extensively studied via autoclave and reactor corroded samples, the initial corrosion mechanism is not fully understood. A Protochipstrade; environmental holder has been utilized to investigate the early stages of corrosion phenomena in Zircaloy-4 inside a transmission electron microscope. In-situ corrosion experiments were conducted at an elevated temperature in a gaseous mixture of 50% oxygen-50% argon at 1atmosphere of pressure. Using procession diffraction, high-resolution imaging and diffraction patterns, the phases present in each sample were determined prior to and following each experiment. Upon oxidation, both tetragonal and monoclinic ZrO₂ were formed and identified, as well as the remaining base metal. Through an improved fundamental understanding of the early stage mechanism of corrosion in these alloys, it may be possible that the corrosion response of these alloys can be improved through both alloying additions and proper grain boundary engineering methods.

Grain boundaries (GBs) of differing crystallographic character are sinks of differing efficiency for radiation-induced defects. We model the atomic structure of GBs in Cu to ascertain what determines their sink efficiencies. We investigate three specific boundaries whose relative sink efficiencies were previously determined experimentally. We find that preparing minimum energy models of these GBs requires changing the number of atoms in the boundary plane. Vacancy formation energies in such minimum energy GBs correlate well with the experimentally determined relative sink efficiencies of these GBs. Our findings suggest that sink efficiencies of non-coherent GBs are determined in part by de-trapping of absorbed vacancies from the GBs.

Understanding the mechanism and characterizing the diffusion behavior through CVD β-silicon carbide (SiC) is crucial to the licensing of the TRISO fuel design.
The thermal diffusion of three key fission products: cesium, europium, and strontium through SiC was investigated using ToF-SIMS through a multi-layer sample design that mimics the structure of the TRISO particle and allows for study of thermal diffusion without introducing any radiation damage to SiC. Diffusion was investigated between 8000C and 13000C for which cesium was the slowest diffuser, strontium the most rapid and europium in between. Preliminary results suggest that in this temperature range cesium exhibits Type C (grain boundary dominant) diffusion kinetics. Strontium exhibits both Type B (competing grain boundary and bulk) diffusion kinetics and Type A (bulk dominant) diffusion kinetics with a transition temperature between 11000C and 12000C. Finally, in this temperature range europium exhibits all three regimes of diffusion kinetics starting with Type C between 8000C and 10000C, Type B in the 11000C and 12000C range, and Type A over 13000C.

We report on uncertainty quantification and sensitivity analysis of Kinetic Monte Carlo (KMC) simulations that examine the behavior of defects in nuclear materials on the mesoscale. Our goal is to examine defects, created by the irradiation of materials that can affect the properties of the material and may cause catastrophic failure of the materials. Our model looks at the diffusive behavior of different types of defects in the materials in order to determine the effect of defect evolution on physical properties of the materials. In particular we look at the evolution of vacancies in fluorite lattice found in oxides commonly used in fuels. In addition due to the lack of experimental data on commonly used fuel materials (uranium oxide, plutonium oxide and thorium oxide) we examine ceria as a surrogate material. This allows us to compare the accuracy of our results with those found in experiment. In general KMC simulations utilize different rates for the potential events that occur. We perform sensitivity analysis of the various rates of events in the KMC rate catalog and determine the various factors that may create uncertainty in the quantification of the final diffusivity results.

Novel and more conventional boron carbides were made into heterojunction diodes with silicon for this study. The boron carbides were based on the cross linking of closo-1,2-dicarbadodecaborane (ortho-carborane; 1,2-B₁₀C₂H₁₂), and cross linking based on the combination and cross-linking of closo-1,2-dicarbadodecaborane (ortho-carborane; 1,2-B₁₀C₂H₁₂) and pyridine. In the latter devices, pyridine concentration was varied; samples with a closo-1,2-dicarbadodecaborane (ortho-carborane; 1,2-B₁₀C₂H₁₂) to pyridine ratio of 1:1 (BC:Py1) and 1:3 (BC:Py3). The result is a nonvolatile robust p-type semiconductor of boron carbide (B₁₀C₂H_x):(C₅NH_x)_y. The I-V characteristic curves for the heterojunction diodes exhibit strong rectification where the normalized reverse bias leakage currents are largely unperturbed with increasing pyridine inclusion. The devices are largely gamma insensitive and yet neutron voltaic properties of these boron carbides is demonstrated. The neutron capture generated pulses from these heterojunction diodes were obtained at zero bias voltage although without the signatures of complete charge collection expected boron neutron capture generated electron hole pair production. These results, nonetheless, suggest that modifications to boron carbide may result in better neutron voltaic materials with linking groups chosen from family of aromatic compounds that stretch between borazine (B₃N₃H₆) and benzene that point the way to a whole family of future studies that may ultimately lead to boron carbides better suited to low power and low flux neutron detection.

The life extension of current pressurized water reactors and the design of reliable next-generation nuclear reactors call for advanced structural steels that can sustain radiation up to several hundred displacements per atom (dpa) at elevated temperatures. Here we performed Fe ion irradiation to 150 dpa at 450 °C on bulk coarse-grained (CG, with a grain size of ~2 mu;m) and ultrafine-grained (UFG, with grain size of ~320 nm) T91 steels [M. Song et al, Acta Materialia, 74 (2014) 285-295]. Extensive microscopy studies show that fine grains in UFG T91 reduced the density of nanocavities and dislocation loops. The swelling rate in UFG steel is three times lower than that of CG T91 due to the existence of abundant defect sinks, such as high angle grain boundaries and dislocations. A strong surface effect with size dependence was noted during heavy ion irradiation studies.

L1₂-ordered γ' precipitates embedded in a fcc γ matrix impart excellent mechanical properties to Ni-base superalloys. However, these precipitates are not stable under irradiation. We study radiation-induced disordering and mixing at a coherent (100) facet of a γ' precipitate neighboring a pure Ni matrix by modeling multiple, sequentially introduced, 10 keV displacement cascades using molecular dynamics. At room temperature, the ordered Ni₃Al disorders rapidly within 0.1-0.2 dpa and then gradually dissolves into the adjacent Ni layer at higher doses. Both the disordering rate and mixing parameter of the ordered precipitate calculated from the simulations are in quantitative agreements with experimental data. The influence of a ternary element upon the precipitate stability is also examined.
This work was supported by the Laboratory Directed Research and Development program at Los Alamos National Laboratory under Project No. 20130118DR, under DOE Contract DE-AC52-06NA253.

Materials with the Apatite structure are considered as constituents for multi compounds nuclear waste forms, due to their compositional flexibility [1]. Therefore, their behavior under irradiation is of significant industrial interest. Molecular dynamics simulations, used in conjunction with a set of classical pair potentials, have been employed to examine simulated radiation damage cascades in the fluorapatite structure. Regions of damage have subsequently been assessed for both their ability to recover and the effect that damage has on the important structural unit namely the phosphate tetrahedron. Damage was considered by identifying how the phosphorous coordination environment changed during a collision cascade. This showed that PO4 units are substantially retained, with only a very small number of under or over coordinated phosphate units being observed, even at peak radiation damage. By comparison the damaged region of the material showed a marked change in the topology of the phosphate polyhedra, which polymerized to form chains up to seven units in length. Significantly, the fluorine channels characteristic of the fluorapatite structure and defined by the structure&’s calcium meta-prisms remained almost entirely intact throughout. This meant that the damaged region could be characterized as an amorphous phosphate chains interlaced with regular features of the original undamaged apatite structure.
[1] R. C. Ewing and L. Wang, “Phosphates as nuclear waste forms” Rev. Min. Geochem., 48, 673-699 (2002).

Zirconium (Zr) alloys are the primary fuel clad materials in most light and heavy water reactors. Creep is considered to be one of the important degradation mechanisms in Zr alloys during reactor operating and repository conditions. Creep, which is a high temperature deformation process under constant stress, is primarily governed by dislocation glide and climb mechanisms in metals and alloys. In the current work, creep behavior of Zr is investigated using discrete dislocation dynamics (DDD) simulations. Both dislocation glide and climb mechanisms are considered in the simulations that proceed under a constant stress ranging from 100 to 200 MPa. The creep rate is then determined from the time derivative of the creep strain. We show that the DDD simulations are able to predict the expected increase in the steady-state creep rate with increasing stress. The creep behavior is complex and anisotropic for HCP Zr with the interactions stemming from prismatic, basal and pyramidal systems. Our DDD simulations, which are performed by activating different slip systems independently, indicate a faster creep rate along the prismatic slip system relative to the basal and pyramidal slip systems as expected in low c/a-ratio materials such as Zr.

The prediction for radiation-induced embrittlement of the nuclear reactor pressure vessel (RPV) steels is critically important for the safe operation of nuclear power plants. Nanoscale Cu-rich precipitate induced or enhanced by neutron irradiation is one of the main origins for embrittlement of RPV steels. Therefore, a detailed understanding of Cu precipitation kinetics is necessary to predict the embrittlement of RPV steels. Thereby, accurate estimation of the diffusivity and solubility limit of Cu is very important.
However, such traditional studies to estimate the basic quantities performed mainly more than 30 years ago, only at much higher temperatures (ge;690°C) than those practically used (~300°C) because of the limited spatial resolutions. Temperatures in excess of the Curie temperature, which is about 760 °C for Fe, cause a paramagnetic-ferromagnetic transition, which cause anomaly in the diffusivity at the transition temperature. Thus, the diffusivity for ferromagnetic Fe and steels at lower temperatures can only be obtained by extrapolation of data acquired in a very limited temperature range from about 690 to 760 °C, or by computer simulation, and more reliable values are strongly required.
In this study, the diffusivity and solubility limit of copper in ferromagnetic iron and an A533B steel were directly measured at lower temperatures (ge;550°C) than in previous studies using a state-of-the-art technique, atom probe tomography (APT). In pure Fe, the diffusivity was determined to be D=0.48 exp(-Qfrasl;(k_B T)) [m²frasl;s] (Q = 3.22 eV), which is, below 650°C, about 1.3 times higher than the extrapolation from previous study considering the magnetic effect. The measured Cu solubility limit was in good agreement with literature. In an A533B steel, for all annealing temperatures, the diffusivity of Cu in A533B steel were found to be 2-3 times higher than that in pure Fe, although the solubility limit of Cu was similar. APT was also used to study the grain boundary (GB) segregation and diffusion. No Cu segregation is observed at GB near the Cu/A533B steel interface.

Ni-Cr alloys play a critical role in a number of structural components in the corrosive environment of light water reactors. While generally viewed as corrosion resistant, these alloys can still be susceptible to intergranular attack and stress corrosion cracking in certain conditions. In this work, we explore the fundamental mechanisms that drive and limit this degradation at the nanoscale by imaging the corroded grain boundaries of both model and commercial alloys with complementary transmission electron microscopy and atom probe tomography. These observations have revealed that the grain boundary diffusivity of Cr at relevant temperatures (330-360°C) is ~100X faster than predicted from literature values at higher temperatures and that diffusion induced grain boundary migration (DIGM) is commonly observed in all Ni-Cr alloys. This results in the rapid formation of prominent Cr-depleted zones beneath the oxidized grain boundary. Increasing the Cr content to 30 at.% effectively stops intergranular oxidation through the formation of a protective layer of Cr₂O₃, while the grain boundary beneath still exhibits the formation of large Cr-depleted zones as a result of DIGM. Analysis of Ni-20Cr, suspected to be near the critical Cr concentration to form protective Cr₂O₃, reveals a far more complicated intergranular oxide morphology that is suggestive of incomplete local passivation.

The development of a uranium alloy with natural corrosion protection is highly desirable to prevent the formation of pyrophoric uranium hydride on contact with water. If such a corrosion resistance could be designed it would be of use both in terms of improving the safety of stored spent fuel as well as a potential alternative to zirconium cladding on fuel pellets. Uranium-niobium alloys are a candidate material for this purpose, but the microstructural mechanism for their corrosion protection is not well understood. Atom Probe Tomography is a microscopy technique where individual atoms are field evaporated from the surfaceof a specimen onto a position-sensitive time-of-flight detector, enabling three-dimensional chemically-resolved atom by atom imaging of the specimen.
A LEAP 3000X HR atom probe instrument was used to study the surface microstructure of a series of air-exposed materials: pure depleted uranium, pure niobium, and alloys of depleted uranium containing 3 wt.% and 6 wt.% niobium respectively. The pure uranium material after several days of air exposure was heavily oxidized, with oxygen content of 25 at.% observed even 100nm deep into the specimen. In samples where air exposure was minimised, a banded structure was observed, where the stoichiometry of the oxide changed from UO2 at the surface to UO deeper into the oxide layer, followed by a pure uranium matrix. In addition, a thin layer of uranium hydride was detected at the metal-oxide interface, indicating that water, rather than oxygen, may be the driver behind oxidation of uranium in air.
A banded oxide layer was also present in both uranium-niobium alloys, consisting of UO2 at the surface and UO directly below. Beneath the oxide, niobium segregation was observed. In the 6 wt.% Nb sample, this segregation formed a continuous layer, with concentrations as high as 65 at.% Nb. Below this layer, the matrix of the 6 wt.% specimens contained concentrations closer to the expected bulk composition, with dramatically reduced oxygen content. The 3 wt.% Nb samples showed a less contiguous array of Nb precipitates interspersed with oxides, with much deeper oxidation. The formation mechanism of this Nb layer and the implications of these microstructural changes on corrosion resistance are discussed.

Solid solution strengthening is one of most widely used methods to achieve desired materials properties, including radiation tolerance, by alloying other elements into pure metals but still remaining entire in solution. The problem of radiation damage in structural materials has been intensively studied for 70 years. While it has long been recognized that specific compositions in binary and some ternary alloys have enhanced radiation resistance, how the structure and chemistry affect defect formation and damage evolution remains unclear, and this has been the standing roadblock to the future-generation energy technologies.
To extend our knowledge beyond current incremental property improvement of traditional alloys that are dominated by a single atomic species with minor secondary alloying element additions, it is essential to gain atomic-level control of local chemical disorder that has a profound impact on defect dynamics in solid solution alloys. Based on Ni-based binary alloys with principal substitution element of Co or Fe, we demonstrate how defect production and damage accumulation under ion irradiation are influenced by modifying chemical complexity and how they can be manipulated to alter radiation performance.
This work was supported by Energy Dissipation to Defect Evolution (EDDE), an Energy Research Frontier Center supported by the U.S. Department of Energy, Basic Energy Sciences.

A new class of material labeled ‘high entropy alloys&’ have been produced by magnetron sputtering and arc melting techniques. These multi-principal component, V and Zr, systems were designed for high temperature phase stability and neutron radiation resistance in the thermal spectrum. X-ray diffraction, electron microscopy characterization before and after ion bombardment was carried out and ab-initio methods were also used to model the material on an atomistic level. Amorphous and disordered body-centered cubic crystal structures were observed. Both single phase and dual phase dendritic microstructures were found with varying compositions.
In addition, these alloys should possess good resistance to radiation damage due to their inherent disorder. It is hypothesized that the high configurational entropy may promote superior point defect recombination compared to other materials. This would make them suitable for use as a coating or fuel cladding in generators 3+.

FeCr steels are promising candidats for structural materials in advanced fission and future fusion reactors. In FeCr alloys where Cr local magnetic moments tend to align anti-ferromagnetically with the moments of both nearest Cr and Fe neighbors in the bcc lattice, magnetic frustrations often occur, particularly close to structural defects and chemical interfaces. In this way, magnetism may strongly influence energetic and mobility of defects. Also, the presence of defects may induce the occurrence of specific magnetic structures aiming at relaxing the frustrations.
Employing Density Functional Theory studies, we have shown that the formation energy and migration barrier in bulk Cr are very sensitive to the underlying magnetic structure [1]. This result is relevant when considering Cr-rich regions in FeCr systems. We have also identified the magnetic frustrations to be responsible for an unusually high formation energy of Fe/Cr(110) interfaces compared to that of Fe/Cr(100). Two ways are suggested to relax partially the magnetic frustrations at the (110) interface and to lower its formation energy. Noncollinear magnetic configurations can be developed where local moments of Fe and Cr atoms are perpendicular to each other. Also, in-plane spin-density waves show a very stable magnetic structure with the nodes at the interface layer. The presence of low-moment sites at Fe/Cr(110) offer another way to relax the magnetic frustrations and lower the interfacial energy [2]. Using effective interaction models coupled with Monte Carlo simulations [3,4], the effect of temperature on magnetic order at interfaces and on interface energies has been investigated. It is found that while the low-temperature noncollinear bulk magnetic configurations of Cr remain stable up to the Neel temperature, the Cr atomic layers close to the interfaces retain their magnetic order well above this temperature.
Finally, in order to gain a closer insight into magnetic properties of precipitates in FeCr systems, emerging during FeCr decomposition for instance, we have studied magnetic structures of various Cr clusters in Fe, as well as Fe clusters in Cr. We have found low-energy non-collinear configurations and non-zero total magnetic moment in the Cr clusters, due to a strong magnetic interaction between the interfacial Cr atoms and the lattice Fe atoms. In addition, due to the interfacial magnetic effect, a significant anti-ferromagnetic correlation may remain in the Cr clusters well beyond the Neel temperature of bcc Cr
[1] R. Soulairol, Chu-Chun Fu and C. Barreteau, Phys. Rev. B, 83, 214103 (2011).
[2] R. Soulairol, Chu-Chun Fu and C. Barreteau, Phys. Rev. B, 84, 155402 (2011).
{3] M. Yu. Lavrentiev, R. Soulairol, Chu Chun Fu, D. Nguyen Manh , S. L. Dudarev, Phys. Rev. B 84, 144203 (2011).
[4] Thanks to a collaboration with Drs. M.Y. Lavrentiev, S.L. Dudarev, and D. Nguyen Manh from CCFE, UK.

In the irradiation environment of nuclear reactors, structural materials are known to form
large concentration of point defects such as vacancies and interstitials by the ballistic
action of fast incoming neutrons. The production and subsequent fate of these point
defects causes multiple pernicious effects in these materials such as radiation induced
hardening and swelling. The potential for such deleterious issues will be much higher
within the structural materials of next-generation fission reactors as these are expected to
withstand a radiation dose about an order of magnitude higher than the current fleet of
reactors. Hence, there is an urgent need to develop the fundamental understanding of the
damage mechanisms produced by these point defects in order to aid in the design of next generation
nuclear materials. One such damage mechanism is radiation-induced
segregation (RIS), which is the micro-chemical segregation of alloying elements to sinks
like grain boundaries caused by preferential coupling of alloying elements with the
radiation-induced defect fluxes to sinks. RIS causes changes in the chemical distribution
within the alloy, modifying the strength and fracture characteristics of the alloy. In this
presentation, first we will present an overview of our recently developed atomistic informed
phase field model to predict solute segregation to sinks, at the continuum level
with minimal experimental inputs. Materials parameters needed at the mesoscale are
obtained from numerical simulations at the atomistic level. Then, we will present our
results on the role of grain boundary character, grain boundary chemistry, grain size,
temperature, radiation dose and dose rate on the segregation of Cr in model Fe-Cr alloys.
The results will be compared with experimental RIS measurements. Finally, based on the
calculated RIS profile, we will present guidelines for designing alloy microstructure that
can minimize RIS.

Stainless steels are used as major construction materials in many important branches of industry such as nuclear power plants, civil industry, and heavy manufacturing industry. Many useful properties of stainless steels are well known to be badly affected by sigma-CrFe. The phase stability of the sigma-CrFe under fast electron irradiation was studied using high voltage electron microscopy (HVEM). Sigma-CrFe transformed into a bcc solid-solution phase under MeV electron irradiation at the temperature range between 298 K and 473 K. In contrast, no changes in the topological structure were observed at temperatures below 103 K. Sigma-CrFe phase became more resistant to the irradiation with decreasing the temperature. The dominant factor affecting the decomposition of sigma-CrFe to solid solution phase under irradiation was discussed in terms of the Gibbs free energy and the microstructural changes associated with the thermal assisted, radiation-enhanced migration of defects and/or constituent atoms. [Reference] S. Anada, T. Nagase, K. Kobayashi, H. Yasuda, H. Mori: Acta Materialia, 71 (2014) 195-205.

This study investigates the mechanisms controlling irradiation induced microstructural evolution, particularly the development of dislocation loops, in ferritic iron chrome alloys over a wide range of temperatures. Cluster dynamics is an expansion of simpler rate theory models which considers the possibility of defect interactions to form arbitrarily large clusters while explicitly including spatially dependent sinks. Primary damage is introduced with a multiscale approach by using a database of molecular dynamics simulations of displacement cascades to implant damage as a distribution of defect clusters rather than as simple Frenkel pairs. The kinetics of the model are built upon observations of defect behavior during in-situ heavy ion irradiation experiments which have shown a drastic increase in defect mobility during irradiation. This model demonstrates the importance of considering such irradiation driven kinetic pathways along with the one dimensional nature of interstitial motion in understanding the development of the microstructure.

In alloys under irradiation, the elimination of excess point defects at sinks, such as grain boundaries, free surfaces or dislocations, can lead to Radiation Induced Segregation (RIS) phenomena. In ferritic steels for example, various cases of Cr depletions or enrichments have been observed at grain boundaries, depending on the alloy composition and microstructure, and on irradiation conditions. We present atomistic Monte Carlo simulations of RIS in binary Fe-Cr alloys, taking into account the diffusion of vacancies and self-interstitials atoms, and using a jump frequency model fitted on ab initio calculations. These simulations are used to measure the Lij coefficients of the Onsager Matrix, especially its non-diagonal terms that control the coupling between of Cr and point defect fluxes, and therefore the enrichment or depletion tendencies. The dependence of the Lijs on the alloy composition and on the temperature are analyzed and related to the migration barriers of point defects. The evolution of segregation profiles are simulated, and compared to available experimental studies. Radiation Induced Precipitation (RIP) in undersaturated Fe-Cr solid solutions are also considered, as well as complex microstructural evolutions resulting from the interaction between segregation and precipitation in supersaturated alloys.

The research for radiation-resistant compounds with fluorite related structure is a challenging area of research for developing reliable nuclear energy systems. Those systems, and advanced fuels in particular, require a sound understanding based on reliable experiments and numerical modelling in order to ensure their safe behaviour throughout the fuel life cycle, from fabrication to end of life storage. The problem of understanding and developing a predictive capability for the evolution of fuels is a challenging one, even for features apparently simple as oxidation. Uranium oxides have deceptively simple approximate chemical formulas that actually betray their extremely complex structural features and their propensity to form nonstoichiometric phases of composition UO_2+x: in these phases U atoms exhibit multiple charge states and configurations. We would try to summarize a multidisciplinary effort to characterize the structural features of laboratory prepared UO_2+x systems (0

Thermal conductivity is a critical property of oxide fuels because it affects both nuclear energy conversion efficiency and nuclear safety. The microstructures in nuclear fuels play important roles in affecting this property. For example, the high density of gas bubbles and grain boundaries in high burnup fuels induces interesting thermal transport properties. In this work, we use phase field modeling to study how the distribution, interaction, and evolution of microstructures affect the thermal conductivity in UO₂. The modeling is conducted using the MARMOT simulation package developed by Idaho National Laboratory, a mesoscale modeling tool based on MOOSE framework. The effects of gas bubbles on the thermal conductivity are studied by varying the gas bubble radius and porosity. The calculation results are compared with analytical models and experiments. Good agreement with analytical modeling and discrepancy with experiments are found. The evolution of thermal conductivity during grain growth is also obtained. The extracted grain boundary Kapitza resistance from simulations is compared with the analytical model and good agreement is obtained. Finally the combined effects of grain boundaries and gas bubbles are studied. It is found that although grain boundaries alone do not affect the thermal conductivity significantly, their ability of aligning gas bubbles at grain boundaries can induce a significant impact

In fission based nuclear reactors, uranium dioxide fuel is subject to an intense neutron environment that drives the fission chain reaction. In this process, fission fragments will be produced with an energy reaching 1 MeV/amu. These fragments will initially lose energy through inelastic interactions resulting in excitations of the electronic structure. The excitations subsequently transfer energy to the atomic lattice through electron-phonon (e-p) coupling resulting in a thermal spike which may enhance mobility of fuel atoms. Consequently, the enhanced mobility resulting from fission energy deposition is expected to promote annealing of lattice defects such as ion tracks. Classical molecular dynamics (MD) simulations of uranium dioxide were performed using the LAMMPS code to investigate the effects of fission enhanced mobility on ion tracks formed in the fuel. The MD model was composed of 10x60x60 unit cells, 432000 atoms, and used a Buckingham potential to describe interatomic interactions. A two-temperature model was used to capture the process of fission energy deposition in the electronic subsystem and its transfer to the atomic lattice through e-p coupling. Previous MD simulations demonstrated that fission-enhanced diffusion became more pronounced as the electronic system behavior was varied from metal-like to insulator-like, i.e., increasing the e-p coupling strength. In the present MD simulations, the annealing of an existing ion track (radius 1.2 nm) due to the interaction with a 18 keV/nm fission fragment was observed. For a metal-like system (weak e-p coupling), it was found that the track persisted with a radius of 0.7 nm. For an insulator-like system (strong e-p coupling), it was found that the track was transformed to disconnected defect clusters each with a radius reaching 0.3 nm.

We present a Monte Carlo solution of the Boltzmann transport equation for phonons in uranium dioxide with various levels of defects. The Boltzmann transport equation is simplified by assigning time scales to each scattering mechanism of phonons. Unlike most other works on solving this equation by Monte Carlo method, the momentum and energy conservation laws for phonon-phonon interactions in uranium dioxide are treated exactly by considering only the interactions that obey the pertinent conservation laws. The simulation scheme accounts for the acoustic and optical branches of the dispersion relationships and considers the 3D representation of the actual shape of the UO₂ Brillouin zone (truncated octahedron) with the assumption of isotropic continuum. Evolution of all polarization branches of phonons in phase space is tracked within the Monte Carlo. Using periodic boundary conditions, our results illustrate the diffusion limit of phonon transport in uranium dioxide, and make possible the prediction of thermal conductivity. A simple kinetic theory model is also implemented in which conductivity is calculated using phonon heat capacity, velocities, and scattering time-scales. The effect of temperature and defect concentration on conductivity is predicted with both models and the results are compared with experimental data available in the literature. This research was supported as a part of the Energy Frontier Research Center for Materials Science of Nuclear Fuel funded by the U.S. Department of Energy, Office of Basic Energy Sciences under award number FWP 1356, through subcontract number 00122223 at Purdue University.

We present results of a comprehensive first principle density-functional theory-based study of phonon and thermophysical properties of the strongly correlated UO₂ system, and compare with results of recent experimental studies. Phonon properties will include the phonon density of states, phonon dispersion curves, Debye temperature, and phonon relaxation time. Thermophysical properties will cover thermal expansion coefficient, free energy, entropy, specific heat capacities of C_p and C_v, mode Gruneisen parameters and thermal conductivity, from 10 to 2000 K. The spin effect will also be discussed. This work is supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program funded by the U.S. Department of Energy, Office of Nuclear Energy.

The semiconductor cubic boron phosphide (BP) (E_g = 2.0 eV) has a high electron mobility (reported to be >100 cm²/Vmiddot;s) and a large thermal neutron capture cross-section of 3840 barns for the ¹⁰B isotope, making it potentially useful for neutron detection. For effective neutron detection, the BP must have a high crystal quality, a low defect density, and low residual impurities to realize the best possible electronic properties. In this work, the structural properties of BP films were improved over direct growth on silicon substrates by using 3C-SiC/Si(100) substrate. 3C-silicon carbide has a much smaller lattice constant mismatch (-4.1%) than silicon (16.4%), an almost identical coefficient of thermal expansion coefficient as BP, and excellent thermal stability thereby minimizing the substrate induced defects in the BP films. The 3C-SiC layer on Si substrate prevents the diffusion of Si atoms into the BP, and the diffusion of boron and phosphorus into the silicon, thereby reducing cross-contamination as is typical with BP deposited directly on silicon. 3C-SiC has the same crystal structure (zinc blende) as BP, thus it eliminates in-plane rotational twinning defects which are seen with BP films on vicinal 4H- and 6H-SiC (0001) substrates, due to a symmetry mismatch. High quality BP films were deposited using PH₃+B₂H₆+H₂ system at 1000^oC-1100^oC and atmospheric pressure. Growth parameters such as temperature, reactant flow rates, film thickness, post reaction cooling rate were evaluated to obtain defect-free single crystalline films. Optical and scanning electron microscopy revealed smooth morphologies with predominantly BP(100) orientation. Raman spectroscopy showed an intense, sharp phonon peak located at around 825 cm^-1; peaks from other boron compounds were absent. XRD revealed the BP films were single crystalline with peak height ratios of BP(100) to BP(220) of 2650 on 3C-SiC/Si. In contrast to BP on on-axis 4H- or 6H-SiC(0001), rotational twins were absent, as determined by x-ray topography. These results suggest that 3C-SiC/Si(100) are suitable substrates for BP electronic device development.

Tristructural-Isotropic (TRISO) coated fuel particles are a type of micro-fuel that is to be used in the next generation Very High Temperature Reactors (VHTRs). In the current TRISO design, the Silicon Carbide (SiC) layer provides the primary barrier against escape of fission products. These particles, however, have been reported to release undesirable metallic fission products such as radioactive silver (^110mAg). The release rate and mechanism of Ag transport through the SiC remains unsolved. One hypothesis for the mechanism of Ag transport through SiC is diffusion along grain boundaries (GBs), which hypothesis is supported by a number of recent studies demonstrating diffusion of Ag is significantly faster in polycrystalline than single crystal SiC [1-3]. In this work we therefore focus on elucidating the Ag transport behavior in grain boundaries, including predicting both the diffusion of Ag in High Energy Grain Boundaries (HEGB) and the microstructure dependence of effective diffusivity in a grain boundary network with GB dependent diffusion coefficients.
The Ag diffusion in HEGBs is simulated using an ab initio based kinetic Monte Carlo (kMC) model. The HEGB is modeled as an amorphous SiC (a-SiC) region and Density Functional Theory (DFT) is used to evaluate the formation and migration energies for a number of local environments. Then the Ag diffusion is simulated with a kMC model using migration barriers drawn from a distribution fit to the DFT data. The Ag diffusion coefficient (D) exhibits Arrhenius dependence and D in the HEGB is predicted to be significantly faster than values previously predicted for bulk and a sum;3-GB. However, the remaining discrepancies between GB diffusion predictions and integral release measurements suggest that other conditions are possibly responsible for fast release of Ag (e.g., radiation enhanced diffusion).
To better understand how effective Ag diffusion might be impacted by microstructure the effective diffusivity in grain boundary networks of general polycrystalline materials is evaluated using a kMC model. The effects of multiple GB diffusivities, grain size, and two- vs. three-dimensional networks are examined. It is shown that the effective diffusivity does not depend on the grain size when GB diffusion is the dominant diffusion mechanism. We find that the behavior of the effective diffusivity is qualitatively the same for two- and three- dimensional models. In addition, we find that the effective diffusivity exhibits large fluctuations due to its dependence on the GB distributions, and therefore the details of the materials microstructure can significantly impact measuring the effective diffusivity in a specific finite-size sample.
[1] Friedland, E. et al. 2009. J. Nucl. Mater. 389: 326-331.
[2] Shrader, D. et al. 2011. J. Nucl. Mater. 408: 257-271.
[3] Khalil, S. et al. 2011. Phys Rev B 84: 214104.

Transition metal carbides combine some ceramics physicochemical characteristics and metals electronic properties. These carbides have a real potential for nuclear environment, in particular titanium carbide (TiC), whose thermal conductivity is partially electronic and increases with temperature. In future nuclear reactors, structural and cladding materials will be exposed to extreme conditions of temperature and irradiation as well as oxidising environment. TiC pellets with several microstructures (grain size between hundred nanometers to several micrometers) were prepared using Spark Plasma Sintering (SPS). Then their oxidation behaviour was studied trough thermal treatments performed at 1000°C for several hours and under low oxygen partial pressure (OPP). Prior to this oxidation treatment, TiC samples were implanted with 800 keV 129Xe++ ions at high fluence (6e16 at.cm-2) to simulate gaseous fission product presence near the surface. The projected range (Rp) was 160 nm and the maximal concentration was about 4%. These parameters were simulated by SRIM software. No oxidation was observed for treatment with OPP < 1e-6 mbar. At higher OPP (= 2e-6 mbar), an oxide layer corresponding to titanium oxides and oxycarbides was formed on the surface which thickness depends on the microstructure. The composition, thickness, growth and morphology of the oxide layer were determined by several characterization techniques (Rutherford Backscattering Spectrometry, Nuclear Backscattering Spectrometry, X-Ray Diffraction, Scanning Electron Microscopyhellip;). Correlations between the initial microstructure and the oxide layer properties (thickness, morphology) were also made. It appears that the microstructure with submicron grains and the highest initial porosity exhibits the larger oxide scale. A global transport of xenon correlated with the oxide thickness was observed for OPP > 1e-6 mbar. Despite a most important oxidation, xenon is globally less released from the microstructure with submicron grains.

The expansion and commercialization of Generation IV reactor designs will require new suitable high performance materials. Specifically, the harsh core environment of these new designs makes material selection very challenging with ceramics among the few candidates providing the desired properties. Silicon carbide is among one of these materials that is very attractive for structural and other applications. However the use of silicon carbide is limited due to difficult processing and fabrication. Here, based on polycarbosilane, a low temperature reaction sintering process for SiC is demonstrated. The microstructure, mechanical and thermophysical properties of SiC sintered bodies, at temperatures as low as 930°C, produced through this route will be presented. Furthermore, the ability of this preceramic polymer for joining of sintered SiC bodies will be discussed.

Silicon oxycarbides (SiC_xO_2(1-x)) are a promising class of radiation-resistant and thermally stable amorphous solids. We present atomic-scale investigations of the structure of these materials using classical potentials and density functional theory. Atomic structures of SiC_xO_2(1-x) of varying stoichiometry are generated via a two-step procedure: amorphous SiO₂ is created first via the classical melt-quench approach and C atoms are subsequently introduced as dopants. We find tendency for C atoms to aggregate. Comparisons with experiments are performed, including the densities, the elastic moduli, and the compositions of SiC_xO_4-x tetrahedral structure units, for range of C concentrations.
This work was funded by the DOE Office of Nuclear Energy, Nuclear Energy Enabling Technologies, Reactor Materials program, under contract No. DE-NE0000533.

Uranium-zirconium (U-Zr) alloys are a potential candidate for transmutation reactor fuels. These fuels can be utilized to burn long-lived minor actinides and fission products in fast spectrum reactors. Metallic fuels exhibit a number of favorable characteristics, such as high thermal conductivity, ease of fabrication, and high fissile density. Previously, most research was focused on fuel performance based criteria, such as the maximum burn-up the fuel can achieve. However, low fluence irradiations are also important since they can provide crucial simulation data for the early stages of microstructural evolution in fuels, as well as a more complete understanding of the fuel behavior. This research investigates a U-Zr alloy which was irradiated to 0.003 and 0.03 dpa at 690^oC in the hydraulic rabbit shuttle in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. A number of post-irradiation examination results will be discussed, including synchrotron x-ray diffraction, pair distribution function, positron annihilation spectroscopy, and electron microscopy.

Understanding of fuel-cladding interaction is critical for evaluation of fuel performance in a reactor environment. Exposure to irradiation and high temperatures results in fuel-cladding interaction, interdiffusion, and formation of undesirable brittle or low-melting point phases, which affect integrity of both fuel and cladding. In this contribution we report the results from ongoing work on characterization of intermetallic phases formed as a consequence of interdiffusion between fuel constituents and cladding in various diffusion couples. The complex fuel-cladding chemical interaction (FCCI) occurring between U, Pu-based nuclear fuel alloys and Fe-based cladding at elevated temperatures has been investigated using selective area diffraction (SAD) and X-ray dispersive spectroscopy techniques in scanning, transmission, and scanning transmission electron microscopes (SEM/TEM/STEM). The discussion on phase evolution will be based on phase segregation mechanisms and equilibrium phase diagrams.
This work was supported by the Fuel Cycle Research and Development (FCRD) program of US Department of Energy.

Fluorite-structured ThO₂ is receiving significant interest as a nuclear fuel material due to its greater abundance than urania, enhanced proliferation resistance, and reduced production of Pu and minor actinides. In practice, UO₂ would be mixed with ThO₂ to sustain fission reactions, and therefore understanding defect dynamics in the solid solution are of great importance. In this work, we study the energetics of formation and migration of oxygen vacancies in ThO₂ based solid solutions, i.e., with CeO₂, UO₂shy;, and PuO₂ from static calculations using pair potentials and from density functional theory calculations. In ThO₂-UO₂ solid solution, we find that addition of UO₂ into ThO₂ greatly decreases the oxygen vacancy formation and migration energies. Among the range of UO₂-ThO₂ solid solutions studies, UO₂ exhibits the lowest formation energy and Th_0.25U_0.75O₂ exhibits the lowest migration energies. Similar result in migration energy is observed for Th_0.25Ce_0.75O₂. However, the formation energy is lowest for Th_0.5Ce0_0.5O₂. Moreover, similar comparisons with ThO₂-PuO₂ solid solutions are performed to draw comprehensive defect energetics among these materials.
This work was supported by the Materials Science of Actinides, an Energy Research Frontier Center supported by the U.S. Department of Energy, Basic Energy Sciences.

A monolithic fuel type is being developed to convert US high performance research and test reactors such as Advanced Test Reactor (ATR) at Idaho National Laboratory from highly enriched uranium (HEU) to low-enriched uranium (LEU). The interaction between the cladding and the U-Mo fuel meat during fuel fabrication and irradiation is known to have negative impacts on fuel performance, such as mechanical integrity and dimensional stability. In order to eliminate/minimize the direct interaction between cladding and fuel meat, a thin zirconium diffusion barrier was introduced between the cladding and U-Mo fuel meat through a co-rolling process. A complex interface between the zirconium and U-Mo was developed during the co-rolling process. A predictable interface between zirconium and U-Mo is critical to achieve good fuel performance since the interfaces can be the weakest link in the monolithic fuel system. A post co-rolling annealing treatment is expected to create a well-controlled interface between zirconium and U-Mo. A systematic study utilizing post co-rolling annealing treatment has been carried out. Based on microscopy results, the impacts of the annealing treatment on the interface between zirconium and U-Mo will be presented and an optima annealing treatment schedule will be suggested. The effects of the annealing treatment on the fuel performance will also be discussed.

During transient regime, the power in pressurized water reactor increases strongly in a very short time. Important stresses are generated within uranium dioxide (UO₂) fuel pellets that may lead to their visco-plastic deformation via dislocation movement.
In the literature, only few studies report on the characterization of the mechanical behavior of polycrystalline UO₂ ( [1], [2]): most of the works were performed on single crystals [3]. Existing codes developed for simulating the behavior of fuel pellets under in-reactor irradiation, are usually based on empirical creep laws. The goal of this work is to contribute to a more physical description of such mechanical phenomena. This requires a detailed experimental characterization of the extended defects created in UO₂ under mechanical stress.
Fresh sintered pellets of UO₂ were tested in compression at a constant strain rate (10^-4 s^-1). These experiments were performed at 1500°C, above the fragile/ductile transition that occurs for UO₂ at about 1000 °C [4], and under a reducing atmosphere (Ar 5%H₂) to maintain the sample stoichiometry. The microstructure of strained polycrystalline UO₂ pellets is investigated at the nanoscopic scale by transmission electron microscopy (200 kV, JEOL 2010) for two different final strain values (8 % and 20 %). This study focuses on the characterization of extended defects: the Burgers vectors, dislocation type and glide system have been defined for each dislocation. This enabled the proposition of mechanism regarding their mobility. At 8% strain, the dislocations are not homogeneously distributed. Most of them are found near grain boundaries with densities varying from one grain to another. This suggests an influence of the crystallographic orientations of grain with respect to the compression axis and a possible influence of neighboring grains. At 20% strain, the dislocation density increases significantly: it is twice as high as for pellets with 8 % strain. Moreover the dislocation distribution is homogeneous over the whole sample.
This communication will be dedicated to the description of these experimental results and to their comparison with literature data ( [1], [2]).
[1] Dherbey, F., Louchet, F., Mocellin, A., Leclercq, S. Acta. Mater. 50, 1495-1505 (2002).
[2] Yust, C.S., Roberts, J.T.A., J. Nucl. Mater. 48, 317-329 (1973).
[3] Byron, J.F., J. Nucl. Mater. 28, 110-114 (1968).
[4] Guerin, Y., J. Nucl. Mater. 56, 61-75 (1975).

Uranium-plutonium mixed oxides with high plutonium content are considered as candidates for fuel used in the IV^th generation of nuclear reactors. The ternary system U-Pu-O has been extensively studied for a range of compositions and at various temperatures, but still, data on some domains remains scarce. It has been shown in previous studies that a miscibility gap exists in the hypo-stoichiometric region UO₂-PuO₂-Pu₂O₃ with one phase poor in oxygen, and the other with an O/M (Oxygen to Metal ratio) close to 2.00. Data on the evolution of this region in temperature, especially in the vicinity of the oxygen content corresponding to the highest temperature at which the gap can be observed, is scarce but crucial for developing advanced fuel elaboration methods, better understanding and modelling of long term storage, behavior under accident conditions, fuel-cladding interactions and predicting fuel restructuring. A high temperature X-ray diffractometer with a dedicated gas control setup was used to observe the described region in-situ.
A mixed uranium-plutonium dioxide containing 46 mol% of plutonium was chosen, as this composition has been subject of numerous studies in the past and a lot of data is already available. The experimental procedure was developed to allow for an isothermal continuous in-situ observation of the sample at elevated temperatures while the two cubic phases were present. The gas used was He/5%H₂ with about 15 ppm of water vapor. The initially stoichiometric sample was first heated to 1500°C under the reducing atmosphere to allow quick reduction. The sample was soaked at this temperature for 6 hours and then rapidly cooled to a chosen lower temperature: 500, 450 and 400°C. During the isothermal plateaus at these three temperatures lasting 10, 15 and 20 hours respectively, XRD patterns were recorded continuously, each lasting about 20 minutes. After fast cooling from 1500°C, the sample was reduced and underwent slow oxidation, traversing the miscibility gap to reach a higher O/M corresponding to the oxygen potential of the gas at the lower temperature.
We have observed reflections of the two cubic phases, with one increasing and the other decreasing in intensity during the plateaus. Based on literature data on the lattice parameter evolution in temperature and O/M, we have estimated the O/M evolution of our samples from the calculated lattice parameters and phase fractions obtained from Rietveld refinement of the patterns. We will present a detailed comparison of this data with values calculated using the Calphad method with a recent version of the FuelBase database. To our knowledge, this is a first study of this type, using a state of the art X-ray diffractometer to observe in-situ changes in the crystallographic structure during oxidation of hypo-stoichiometric mixed oxide fuel materials at elevated temperatures.

The current generation of nuclear power reactors are predominantly light water cooled and uranium oxide fueled. Although very economical, predictable and reliable, a new generation of nuclear fuel is required to improve safety and reliability further. The new fuel development comes with a serious set of challenges: above all they are expected to be as economically viable as the previous generation, whilst improving all other aspects of material property relevant to reactor fuel.
Uranium silicides are now being considered for use as a fuel, often incorporating nitrides as a ceramic composite. This deviation from oxide fuel provides a material with a higher thermal conductivity and better behaviour in the event that water comes into contact with the fuel. The silicides come with their own set of modelling challenges, discussed in this work, and for full reactor deployment they need to be met.
The work to be presented will show some of the combined global effort that has been carried out so far in the Westinghouse led CARAT program on accident tolerant fuels, which ANSTO is a part of. This combined theoretical-experimental effort has been shown to provide confidence in materials behaviour whilst paving the way for fuel code development that is required for these new fuels to safely be used in the reactor. Focus will be given to the U-Si system and how to model it reliably and some of the preliminary results attained regarding its expected behaviour in reactor.

A case basis for the production of an ‘engineered nuclear fuel&’ was previously made with the observation that fission product iodine from high burn-up LWR/BWR fuel sequesters in the epsilon (ε) phase, a five metal alloy primarily comprised of Mo-Tc-Ru-Rh-Pd that agglomerates in low-energy sites of uranium grain boundaries during in-reactor operation. Evidence for the formation of silver-iodide (AgI), post ammonium carbonate peroxide (ACP) fuel recycle, was found in the Pd rich portion of the #603;-phase that incorporates silver as a silver-palladium alloy after separation from the primary fuel matrix. Nitric acid recovered #603;-metals have been shown to essentially be devoid of iodine(~3 wt%) while the ACP method preserved iodine ( ~48 wt%) on the high surface area Ag-Pd alloys from aqueous iodine. Efforts to leverage this observation has led to the production of various forms of uranium oxide (UO₃/U₃O₈) fuel precursors that retain various concentrations of silver by direct recrystalization of the uranyl nitrate and silver nitrate salts and through the precipitation method using oxalate as an organic complexant for both UO₂²⁺ and Ag⁺.
We present evidence, after fuel recycle using the ACP method, that iodine is retained in the silver enriched portion of the #603;-phase. Using this evidence we further elucidated on a path forward for the general design of an engineered actinide bearing nuclear fuel that has been shown to retain both chlorine (Cl-) and iodine (I-) at treatments up to 900 °C in air for several hours.
Our findings will have impact on future nuclear fuel designs with regards to ¹²⁹I fission product management, with silver acting as a ‘getter&’, for long-term geological storage and front-end nuclear reprocessing strategies for both the radiopharmaceutical industry and commercial spent fuel inventory. This work leverages PNNL&’s patentable rights, internal LDRD funds under NSD, and further exploits the Radiochemical Processing Laboratory&’s (RPL) ability to prepare and characterize these materials.

Using molecular dynamics the thermal expansion, specific heat capacity and oxygen diffusion have been investigated from 300 K to 3600 K for UO₂, ThO₂ and three compositions of the mixed oxide (U,Th)O₂ system. At high temperature a ‘bump&’ is observed in the specific heat and thermal expansion of UO₂ and ThO₂. This is associated with a super-ionic transition on the oxygen sublattice. Our work is consistent with this transition occurring at a lower temperature in UO₂ than ThO₂, additionally the transition temperature is further reduced in certain compositions of mixed oxide. Our analysis of the formation and diffusion of defects on the oxygen sublattice provides a mechanistic explanation for the high temperature behaviour of thermal expansion and specific heat capacity for mixed oxides. Preliminary results on (U,Pu)O₂ will also be presented.

Order/disorder processes between, U/Nd, oxygen and vacancies were studied by X-ray diffraction in various U_1-yNd_yO_2±x compositions. It was found that adjustment of oxygen concentration in U_1-yNd_yO_2±x solutions with different Nd concentration is accompanied by vacancy formation at U/Nd and oxygen sublattices. Lattice parameters in U_1-yNd_yO_2-x deviate from the Vegard&’s law for high concentration of Nd. Lattice parameters and d-spacing consistently decrease with oxygen concentration at all Nd concentrations indicating vacancy formation during oxygen adjustment at high temperatures. Change in vacancy concentration after oxygen adjustment is estimated relative to Nd concentration and oxygen stoichiometry.
Research supported by the US Department of Energy, Office of Nuclear Energy, Fuel Cycle Technologies Program.

Fouling in Nuclear Reactors
The formation of fouling deposits (CRUD) on fuel cladding in pressurized water reactors (PWRs) presents problems for heat transfer, axial power shift, chemistry control, and worker dose. One solution may be to modify the cladding surface to prevent CRUD adsorption. A combination of two approaches, one chemical and one geometrical, may yield the most promising solution to CRUD, and quite possibly to fouling in general. Understanding what makes a surface CRUD resistant should translate well to other areas that suffer from fouling, such as oil & gas pipelines, refineries, nuclear steam generators and condensers, and high power heat exchangers in general.
Chemical Fouling-Resistant Surface Modifications
A CRUD-resistant surface modification would ideally be thin, corrosion and wear resistant, and would not interfere with neutronics or heat transfer. Rather than applying a coating, which could debond or flake off, a microstructurally graded surface modification of ~100 nm is desired. This is thin enough to let neutrons and heat pass through, almost regardless of its composition.
Several candidate materials fabricated by thin film techniques are undergoing a battery of tests to determine the quality and nature of their CRUD resistance. These include well-characterized materials (Au, Ag), prototypical materials (native ZrO₂), and new materials (ZrN, ZrC, ZrB₂ and their Ti-analogues, ceria, diamond like carbon, and graphene). Pool boiling in simulated PWR coolant containing H₃BO₃, LiOH, and nanoparticles of NiO, Fe₃O₄, and NiFe₂O₄, commonly found in PWR CRUD deposits, is being performed to screen materials for fouling resistance. AFM force-spectroscopy (AFM-FS) using CRUD-functionalized cantilevers directly measures the interfacial adhesion strength of the CRUD-clad bond, as a function of material, particle-surface dwell time, and the environment (air, distilled water, PWR water).
Atomistic simulations of the interfacial bond between CRUD constituents and candidate materials are also being performed. Relaxed bulk crystal structures of each pair of materials are joined, surrounded by vacuum boundary conditions, and strain imposed on each interface via shifting the atomic z-positions of the CRUD without relaxation. AFM-FS measurements are directly compared to these simulations of out-of-plane interfacial strain, with the aim of using atomistics in a predictive manner.
Geometric Fouling-Resistant Surface Modifications
CRUD morphology can also be controlled via induced microcavities, which act as bubble nucleation sites. These bubbles eventually form vapor chimneys, allowing boiled coolant to escape, therefore cooling the CRUD and underlying cladding. There exists a critical spacing of microcavities below which there is no further effect observed on chimney formation. A dense, homogenous distribution of vapor chimneys through the CRUD layer can increase heat transfer and improve thermal margins in nuclear plants.

The susceptibility of light water reactor (LWR) zirconium-based clad alloys to oxidation and resultant large enthalpic heat generation and hydrogen production has prompted exploration of alternative materials. Among those being investigated are Fe-Cr-Al based alloys and SiC composites, most likely with a metal sheath to provide hermiticity and protect the SiC from dissolution in coolant. Thermochemical calculations predicting behavior in steam/hydrogen environments, analysis of oxidation kinetics, and compatibility experiments have been performed. Thermomechanical analysis of composite structures has provided an understanding of the response to stress in the SiC composite systems. The results indicate these materials systems will offer significant advantages over current clad alloys, increasing the tolerance to high temperature oxidative environments.
This research was supported by the U. S. Department of Energy, Office of Nuclear Energy Fuel Cycle Technology Program.

As it has been already shown in literature, reduction of grain size down to several tens or hundreds of nanometers leads to the enhancement of radiation resistance of metals [1-4]. Based on this approach, the aim of the Labex EMC³ (Energy Materials and Clean Combustion Center) project “Naninox” is to (1) establish links between the microstructure, the corrosion resistance and the hardness of a nanostructured stainless steel and (2) study the stability of the microstructure under ion irradiation ; thanks to this better irradiation resistance, a better corrosion resistance and higher mechanical properties after irradiation are expected in the ultra-fine grained stainless steel.
Ultra-fine grained 316L austenitic stainless steel samples have been produced by high pressure torsion (HPT) at 430#730;C and then ion irradiated in Jannus facilities (CEA Saclay) at 450°C and 5 displacements per atoms.
Before and after irradiation, the microstructure has been described by using atom probe tomography and transmission electron microscopy. Corrosion behavior in NaCl solution was tested and nano-indentation tests were performed. All results obtained on the ultra-fine grained 316 austenitic steel are compared to those obtained on the classical 316 austenitic steel in order to discuss advantages of the grain refinement.
References:
[1] B. Radiguet et al. J. Mater. Science 43 (2008) 7338-7343
[2] N. Nita et al. J. Nucl. Materials 329-333 (2004) 953
[3] M .Samaras et al. Phys. Rev. LEtt 88 (2002) 12
[4] W. Voegeli et al. Nucl. Instrum. Methods B 202 (2003) 230

Low temperature hardening and loss of ductility under irradiation are one of the limitations of HT-9 steels in nuclear applications. Formation of second phase precipitates such as α&’ are one of the important factors causing the degradation of mechanical properties. This research focuses on the formation of α&’ precipitates and their effect to the microstructure and mechanical properties of HT-9 steels. Some previous studies suggest that α&’ nucleates on dislocation loops at low temperatures (below 450 °C) during irradiation. In order to understand the effect of the α&’ on the mechanical properties, HT-9 samples were shocked to create a high density of dislocations at a peak pressure of 11 GPa and subsequently annealed at 380 °C and 440 °C for 18 months in an attempt to enhance α&’ nucleation. The mechanical properties of as shocked together with shocked and annealed samples were compared using nanohardness, microhardness and compression tests. Compared to the other samples, shocked and 380°C annealed sample exhibited considerably high hardness which is presumed to be resulted from α&’ formation. X-ray diffraction and transmission electron microscopy based techniques were used to investigate the microstructure. Initial investigations showed that this sample exhibits very fine structure with high dislocation density and twins.