Symposium ES5 : Materials Research and Design for A Nuclear Renaissance

Demands on the properties of structural materials are constantly being increased, new alloys are being developed, and the technology of producing permanent joints for these alloys is crucial for advancement of the nuclear industry, especially for the next generation of nuclear reactors.
One of these developing technologies is high temperature brazing, which has several advantages over welding. The major advantages are the homogeneity of the seam and minimization of the influence of thermal effects on joined materials. Currently, nanocrystalline and amorphous filler metals based on nickel, titanium, zirconium and copper, which have unique properties are widely used. An example is the development of joining technology that uses brazing with these filler metals to connect structural elements of fusion reactors (first wall and divertor). Presently this joining method is being developed for a new generation of fast reactors, where heat-resistant, difficult-to-weld, complex-alloyed steels like NT-9, EP-450, EP-823, ChS-139 (12Cr-Si-Mo-W-Ni-V-Nb-C) are being deployed.
Investigation currently focuses on a rapidly-quenched nickel-base alloy (Ni-Cr-Si-Fe-Mo) in the form of a ribbon that was designed for brazing of these steels with a melting interval 1067-1130^oC. Production technology of an amorphous ribbon has been optimized. Brazing of layouts of fuel elements (connection: plug-tube) of steels ChS-139 and EP-823 (EP-450) has been conducted. It was established that the shear strength of the brazed layout, exceeds the level of allowable stress for the materials of shells and plugs of the fuel elements. All samples have been observed to fail in the steel and not the joint.
The use of nickel-base filler metals is unacceptable for fast lead-cooled reactors such as BREST-OD-300 due to the low corrosion resistance of nickel alloys in lead. Therefore an iron-based filler metal based on the Fe-Cr-Si-B system with a melting interval 1100-1200^oC has been developed, especially for the possibility of making brazed joints of the ferritic-martensitic steel EP-823. The technology of its production in the form of a rapidly-quenched ribbon with amorphous-crystalline structure has been worked out. The optimum brazing mode has also been determined. Strength and corrosion tests, metallographic studies and electron microprobe analysis of the brazed joints have been conducted.
It was established that brazed joints, using iron-base filler metal, have corrosion resistance in lead no worse than that of the base material, and that they can withstand high loads, e.g., tensile strength in uniaxial tension is 750 ± 90 MPa. These results encourage us to recommend the employment of high-temperature brazing using the iron-base rapidly-quenched filler metal as a replacement of currently employed welding techniques, improving the manufacturing technology available for production of new reactor equipment.
The work was supported by the project of improvement the competitiveness of NRNU MEPhI.

Advanced reduced-activation ferritic/martensitic steels and oxide dispersion-strengthened steels exhibit significant radiation embrittlement under low temperature neutron irradiation. Although the phenomenon has been extensively investigated in various materials, including the vessel reactor materials, understanding of embrittlement of f/m steels is unsatisfactory.
Several f\m steels were investigated following neutron irradiation up to ~ 70 dpa. F82H, Eurofer97 and ODS Eurofer were irradiated within the framework of ARBOR projects at 250-340 °C. All the three steels showed large increase in the ductile to brittle transition temperature (DBTT) below 10 dpa. After irradiation to 20-30 dpa the DBTT shift exceeded 200 °C. Ref. [1] suggests that TEM-visible irradiation defects do not explain the observed hardening of irradiated Eurofer97. Embrittlement of ODS steels is even less well understood.
We focus on atom probe tomography (APT) of Eurofer97 and ODS Eurofer steels irradiated with neutrons and heavy ions at low temperature. TEM studies reveal dislocation loops in the neutron-irradiated f\m steels. At the same time APT shows early stages of solid solution decomposition [2]. High density (10²⁴ m^–3) of ~3–5 nm clusters enriched in chromium, manganese, and silicon atoms are found in Eurofer 97 irradiated in BOR-60 reactor to 32dpa at 332°C. In steels, irradiated with Fe ions to the dose of 1.6×10¹⁶ ions/cm², pair correlation functions derived from APT data show Cr-enriched pre-phases. Our APT study of ODS steels irradiated at low temperatures found significant increase in the density of nano-sized clusters. TEM also showed increased density of small (< 5 nm) oxide particles. Nucleation of dislocation loops is suppressed in the regions with high density of nano-oxides.
These results suggest that irradiation-induced nucleation of very small precipitates may represent the main cause of low temperature radiation embrittlement of f\m steels.

1. C. Dethloff, E. Gaganidze, J. Aktaa. Journal of Nuclear Materials 454 (2014) 323–331
2. S.V. Rogozhkin, A.A. Nikitin, A.A. Aleev, A.B. Germanov, A.G. Zaluzhnyi. Inorganic Materials: Applied Research 4 (2013) 112–118.

Strength and ductility of reactor metastable austenitic steels depend on the prior state of the material and are determined by the combined effect of some factors such as radiation and unstable austenitic matrix. Neutron irradiation results in a significant decrease of uniform elongation and increase of material strength. During the deformation at room and low temperatures, additional strengthening can be achieved by forming in the austenite matrix solid plates of α'-martensite. Optimal strengthening of local micro volumes of the material prevents the formation of a stationary neck and increases plasticity. Moreover, second-order effects are those associated with phase stability and consequences of transmutation growing become to first order importance at higher damage levels associated with PWR or VVER plant life extension.
As the object of study, several American (AISI 304, AISI 321, and AISI 316) and the Russian 12Cr18Ni10Ti austenitic steels (analogue of AISI 321) were chosen. These steels vary in element composition such as nickel and manganese. Stacking fault energy as one of the defining parameters of the stability of the austenitic matrix was calculated.
For mechanical testing flat (dimensions of workspace – 7.62×1.52×0.76 mm) and cylindrical (working length - 10 mm, diameter - 1.7 mm) steel samples were machined. Some samples were austenized (1050 °C, 30 minutes) and then irradiated in the core of the WWR-K research reactor at a temperature of about 80 °C to a maximum neutron flux of 1.3x10²⁰ n / cm² (E> 0,1 MeV). WWR-K is 6MWt Kazakhstan pool type research reactor on the thermal neutrons.
Mechanical testing was performed with strain rate of 0.5 mm / min at room temperature. Simultaneously amount of α'-martensitic phase induced in a sample and its geometric dimensions were registrated.
As a result of experiments mechanical and energy characteristics of the materials were calculated. The influence of the chemical composition, structure and neutron radiation on the strength and ductility of austenitic steels were determined. Accumulation diagrams of martensitic α'-phase were built and the parameters characterizing the kinetics of martensitic transformation were calculated.
It is shown that neutron irradiation lowers the ductility and increases the strength of all test materials. Adding nitrogen and manganese provides a high level of strength characteristics of the steel by solid solution hardening and a decrease in ductility.
It was revealed that the martensitic transformation is much more difficult with increasing content of nickel and manganese in the matrix. The neutron irradiation shifts the beginning of martensitic transformation to lower deformation values.

A good understanding of structure-property correlations in neutron irradiated advanced steels is indispensable for design, lifetime prediction and economical operation of fast breeder and future fusion power reactors. Irradiation campaigns on ferritic-martensitic 9CrlWTaV steels (mostly EUROFER) have been performed at the HFR reactor in The Netherlands up to an accumulated damage dose of 16.3 dpa at irradiation temperatures between 250 and 450°C. In the first campaign 80-5100 appm He has been produced by ¹⁰B doping to analyze after irradiation systematically Helium embrittlement at low and high strain rates. While tensile and fatigue tests at low stain rates show no Helium effect up to ~400 appm, Charpy tests at high strain rates reveal a helium related pronounced increase in DBTT of more than 100°C already at 80 appm He. In the second campaign 50 kg EUROFER-ODS was fabricated by powder metallurgy. Compared to EUROFER and other 9CrlWTaV steels the ODS-steel shows in the entire temperature range convincing tensile and fatigue properties after neutron irradiation: much less irradiation hardening, unprecedented uniform elongation and no dislocation channeling induced strain localization. It is to our knowledge the first time that results on neutron irradiated ODS-steels are presented in such a wide temperature range. Correlations are made between strengthening, plasticity, cyclic softening, fatigue lifetime and fracture behavior on the one hand and microstructure evolution on the other hand. Finally conclusions are drawn regarding the microstructural engineering of irradiation tolerant materials.

Zirconium-alloys have been used successfully as materials for fuel claddings in Light Water Reactors (LWR) for many decades. However, in a loss of coolant event, the rise in temperature can result in severe exothermic oxidation of zirconium alloy in conjunction with hydrogen generation from reaction between zirconium and steam. This scenario was tragically borne out during the Fukushima accident in Japan in 2011. In response to this accident, efforts are underway world-wide to develop LWR fuel claddings that are more resistant to oxidation at high temperatures. Approaches being investigated include total replacement of zirconium-alloy cladding with an FeCrAl alloy, molybdenum, and SiC-SiC composite or deposition of oxidation coatings (e.g., Cr, FeCrAl, Ti-Al-C MAX phase compound, and transition metal silicides) on the present zirconium-alloy. Each approach presents its own set of opportunities and challenges that involve considerations of oxidation resistance, high temperature mechanical behavior, manufacturability, and neutronics. The presentation will highlight the advances in these various approaches to the development of accident tolerant fuel cladding.

CRUD, named for Chalk River Unidentified Deposits, is a hydrophilic porous layer that collects on the surface of nuclear fuel pins during normal operation. It is composed primarily of nickel oxide (NiO), iron oxide (Fe₃O₄), and trevorite (NiFe₂O₄). These oxides are corrosion products of the reactor coolant system that are transported as particles throughout the reactor primary loop. As nucleate boiling occurs on fuel rods, the oxide particles precipitate forming CRUD onto the fuel pins. CRUD can grow to 10-100 μm thick with average pore size, porosity, and roughness values between 0.1-1 μm, 40-50%, and 0.5-3.0 μm, respectively. CRUD also contains characteristic boiling chimneys, which are formed as water is pulled into the porous layer through capillary wicking, evaporates and escapes through the chimneys back into the coolant flow.
Although CRUD effects on corrosion and reactivity have been studied extensively, its effects on reactor thermal-hydraulic performance are not well understood. Typically, CRUD buildup is thought to be a purely negative by-product of harsh nuclear environments that increases fuel’s effective thermal resistance and temperatures. However, CRUD is a naturally occurring hydrophilic porous layer that, in surface engineering studies, has been found to enhance CHF and increase the Leidenfrost temperature. Therefore, in thin layers, CRUD has the potential to postpone the boiling crisis in loss of flow accidents (LOFAs) or transient overpower, and accelerate quenching heat transfer following a loss of coolant accidents (LOCA).
In this work, we present the results of a recent investigation of such effects conducted in the MIT Nuclear Science and Engineering department. The boiling performances of CRUD’ed surfaces have been investigated by preparing synthetic CRUD on indium tin oxide-sapphire heaters and testing them in pool boiling and flow boiling conditions. Reactor CRUD was simulated using layer-by-layer deposition of 100 nm silica nanoparticles to form carefully engineered porous, hydrophilic thick films. Photolithography was used to manufacture posts that were then dissolved to create chimneys characteristic of actual CRUD. Synthetic CRUD thickness, wettability, pore size, and chimney diameter and pitch were verified to be representative of reactor CRUD. During testing, IR thermography and high-speed video have been used to shed light on the boiling mechanisms. Measurements enabled by such diagnostics include: time dependent distributions of temperature and heat fluxes on the boiling surface, typically with a <100 µm space resolution and ~1 ms time resolution, nucleation site density, bubble departure frequency and diameter.
Pool boiling experiments have revealed that heaters with synthetic CRUD may have up to 100% enhancement in CHF, while minor effects were experienced in the heat transfer coefficient in both pool boiling and flow boiling conditions. CHF in flow boiling conditions is currently being investigated.

The development of new cladding materials for light water reactors (LWR) has been seriously considered with improved high temperature oxidation resistance to minimize hydrogen production from an accident such as a Loss of Coolant Accident. FeCrAl alloy has been selected as a candidate material for its improved properties on oxidation resistance over the present LWR zircaloy cladding. Because of higher thermal neutron absorption for FeCrAl alloy, the cladding wall thickness has to be reduced down to ~ 350 µm that is significantly thinner than that of the current zircaloy cladding in LWR. New weld techniques of joining the endplug to cladding need to be developed. This presentation will summarize the recent work on the weld development for thin-walled FeCrAl cladding. The microstructural characterization of the weld-affected zone and mechanical property investigation of the weld zone will be discussed.

This work has been focused on joining the emulated FeCrAl thin-walled cladding to the end plug. Both laser beam weld and pressure resistance weld are investigated. The former is a fusion-based weld technique with high intensity pulsed laser to minimize the heat affected zone while the latter is a solid state joining without melting to avoid significant change to the original microstructure. For laser weld, it requires optimization to ensure sufficient heat penetration with minimum material loss from ablation. The pressure resistance weld requires securing the contact through breaking the surface oxide and small amount of deformation followed by a rapid and uniform heating at the contact with minimum additional deformation on the cladding. Both techniques show promising results based on the tensile tests. The characterization of the weld including X-ray 3D tomography for nondestructive examination, optical microscopy and SEM for cross-section samples, tensile test for the joined cladding-endplug set, micro hardness test for weld zone of the cross-section samples, EBSD analysis for grain texture, and hydraulic pressure burst test of the jointed cladding-endplug set. Post-weld heat treatment reveals significant improvement on mechanical property of the weld.

The particulate fouling of heat exchanger and pipe surfaces in operational power plants is problematic for continued efficient operation. Fouling deposits on the fuel of nuclear reactors, referred to as CRUD, cause additional operational concerns such as axial power shifts and accelerated corrosion. The adhesion of fouling deposits to surfaces is partially defined by the attractive van der Waals (VDW) force, the magnitude of which is given by the Hamaker constant. Therefore, developing materials with minimized Hamaker constants should minimize VDW bonding, subsequently reducing or eliminating fouling at its source.
The Hamaker constant is defined by the strength of electronic interactions between two solids through an intermediating fluid. Theoretically, the interaction of any of the two solids with the mediating fluid can define the magnitude of the Hamaker constant. This magnitude is defined by the similarities of the dielectric spectra of each material and the fluid, which can be approximated by their visible indices of refraction. Therefore, developing a material which dielectrically mimics its surrounding fluid should theoretically lead to a Hamaker constant of zero, no VDW attraction, and reduced fouling.
In the present study, we describe theoretical, computational, and experimental efforts to design material surfaces which mimic high temperature, high pressure water for fouling resistance in nuclear and geothermal plants. The VDW force that enables fouling particles to adhere to surfaces is measured using atomic force microscope force spectroscopy (AFM-FS), and coatings that are expected to minimize this force while retaining stability at operating conditions are suggested. We focus on the fluorine-containing family of minerals, whose calculated Hamaker constants are two orders of magnitude lower than those of the oxides found in nuclear and geothermal energy systems. The theoretical prediction of the adhesive force is compared to AFM-FS measurements in water, nitrogen, and vacuum. The VDW force is shown to be calculable directly from density functional theory (DFT) simulations of materials, opening the way to computational discovery of new fouling-resistant materials by a genetic algorithm approach.

Tungsten is a candidate material for the divertor and for first wall armour of future thermonuclear fusion reactors (ITER and DEMO). In the near surface of plasma facing materials, high densities of interstitials and vacancies are produced in addition to the implantation of high concentrations of helium as well as hydrogen isotopes from the plasma. These defects and gases will induce changes in the microstructure and can lead to dramatic changes in materials properties. In this perspective, the fate of irradiation induced defects and their interactions with helium atoms and hydrogen isotopes has to be understood. We use a combination of modeling techniques, namely density functional theory calculations, molecular dynamics and object kinetic Monte Carlo methods, which span different time and length scales to tackle the problem. In this work we show how modeling and carefully designed experimental investigations can provide a route to the understanding of the microstructure evolution of materials in these conditions. In particular, we will address self trapping trap mutation for helium atoms and the formation of helium bubbles, the formation of nanovoids in the near surface region of 800 keV ³He implanted tungsten, and the trapping of H atoms by self interstitials and dislocation loops.


This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

Oxide nanoceramics combine the enhanced radiation tolerance of nanocrystalline materials with the chemical inertness of oxides, and are promising coating materials for highly corrosive and intense radiation environments. In this work, the properties of oxide nanoceramic coatings are evaluated as radiation damage approaches extreme levels, reaching or even exceeding those anticipated for advanced nuclear systems. In particular, taking Al₂O₃ as a model coating material, an effort is made to establish a correlation between the irradiation spectrum and the observed evolution of the oxide’s structural features and mechanical properties. The irradiation conditions are relevant for advanced nuclear systems, and consist in low, moderate, high and extreme radiation damage levels at 600°C –namely, 20, 40, 150, 250 and 450 displacements per atom. A comprehensive analysis of the irradiated nanoceramic is accomplished by X-Ray Diffractometry (XRD), Transmission Electron Microscopy (TEM), Scanning-TEM (STEM), and nanoindentation. The results confirm that grain growth is the main structural change induced by irradiation in oxide nanoceramics. This structural change manifests mechanically through an initial increase of hardness that is well fitted by the Hall-Petch relationship, and eventually through softening in the extreme damage range. Stiffness increases sub-linearly with damage before reaching a plateau. Further, both hardness and stiffness depend on the phase present. The phase evolution may be depth-dependent, and depends strongly on the ion utilized and on the irradiation spectrum. To conclude, it is shown that the coatings successfully provide negligible hydrogen permeation and superior corrosion resistance to structural steels at high temperature in chemically aggressive media. All these features are of particular interest for accident tolerant fuels, Generation IV reactors and fusion systems.

Fuel rod cladding is the first barrier for retention of fission products and the gross structural integrity of the cladding ensures coolable core geometry. To ensure adequate cladding performance during Emergency Core Cooling System re-flooding and during loss-of-coolant-accidents (LOCA), the current NRC licensing criteria limits the peak cladding temperature and the maximum cladding oxidation during the accident. The purpose of these limits is to prevent cladding embrittlement during a LOCA to mitigate the potential for subsequent catastrophic widespread cladding fracture. The purpose of this work is to obtain experimental data on the kinetics of high-temperature oxidation and residual ductility of sponge-based E110 cladding, and provide the data needed to assess the effects of hydrogen pickup on cladding embrittlement during the accident.
Two-sided oxidation experiments were recently conducted at 1000-1200°C in flowing steam with samples of sponge-based Zr-1Nb alloy E110. Although the old electrolytic E110 tubing exhibited a high degree of susceptibility to nodular corrosion and experienced breakaway oxidation rates at relatively low test times, the new sponge-based E110 has been demonstrated as having steam oxidation behavior comparable to Zircaloy-4. The sponge-based E110 followed the parabolic law, and the derived oxidation rate constant is in good agreement with the Cathcart-Pawel (CP) correlation at 1100-1200°C. For 1000°C oxidation, the weight-gain of sponge-based E110 is much lower than Zircaloy-4. No breakaway oxidation was observed at 1000°C up to 8000s. Ring compression tests were conducted to evaluate residual ductility of oxidized samples at room temperature and at 135°C. All sponge-based E110 were still ductile at 135°C after being oxidized up to 20% equivalent cladding reacted at 1000-1200°C. Microhardness tests, metallographic examinations, and scanning electron microscopy examinations were performed on oxidized E110 specimens to correlate the material performance to their microstructures.
In addition, a post-quench ductility study was conducted at 1200°C with tubing specimens of Zircaloy-4, 347 stainless steels, and the commercial FeCrAl alloy APMT under design basis LOCA and extended LOCA conditions. The parabolic rate constant for Zircaloy-4 tubing was in excellent agreement with the CP correlation. The APMT alloy experienced the slowest oxidation rate among all materials examined in this work. SS-347 was still ductile after being oxidized for 2400 s (CP-ECR=50%), but the maximum load was reduced significantly owing to the metal layer thickness reduction. No ductility decrease was observed for the post-quenched APMT samples oxidized up to four hours.

The nuclear sector is currently undergoing a renaissance with 160 new fission reactors being built and another 300 planned. The move to Gen-IV and fusion reactors aims to improve on current technology by increasing output and efficiency but will entail creating more extreme environments within the reactors putting added demands on materials. To this end, novel functional and structural materials are being developed capable of withstanding such extremes. Functional materials are often designed to have properties which make them highly suitable for a particular task but can lead to complex material properties which are difficult to predict. During fabrication of the materials small variations in parameters (e.g. heating/cooling rates, alloying percentages etc.) can yield drastic differences in material properties. It is therefore easily possible to create a large number of new materials quickly but characterising them so that they’re well understood, and therefore useful, can be a lengthy and costly process.

The performance of some of the functional materials proposed (e.g. foams and composites) is dominated by their complex and often anisotropic micro-features. This paper uses the ‘image-based finite element method’ (IBFEM) to perform virtual characterisation of thermal properties. By this method an X-ray tomography scan of graphite foam is converted into an ultra-high resolution FEM model which captures the micro-features. Samples are first digitally cut and prepared from a parent block. The virtual samples are used in an in silico reproduction of the laser flash laboratory experiment. This enables measurement of the material’s thermal diffusivity. Variation in the results due to orientation of the sample demonstrates the power of IBFEM to capture anisotropy due to micro-features.

This method could be expanded to a full suite of laboratory experiments, paving the way to fast virtual material characterisation which will increase the turnaround of new material development.

Oxide Dispersion Strengthened (ODS) ferritic alloys have been considered as candidate materials for nuclear applications due to their excellent resistance to radiation damage, high temperature strength and creep resistance. However, the application of ODS ferritic alloys are limited by the low formability. The maximum uniform elongations of ODS ferritic alloys under 200°C before necking occurs are limited to 5-7%, which poses a significant challenge in shaping these alloys for nuclear reactor applications. Conventionally, to deform such high-strength materials with limit ductility, it requires the workpiece to be heated into the temperature region for either warm forming or hot forming. The thermal softening effect decreases the yield strength of the materials and increases ductility. However, such high temperature can decrease the effectiveness of tools and dies, and is generally very expensive and inefficient.

Here, we propose a new idea to deform ODS ferritic alloys, i.e., applying electric current through workpieces during the plastic deformation, which is known as Electrically-Assisted (EA) forming. In EA forming, constant or pulsed electric current passes through the workpiece during the deformation process to increase the formability of high-strength materials that are difficult to be deformed using conventional forming techniques.

We experimentally investigate the mechanical behavior of ODS 14YWT steel subject to EA deformation. The results are compared with conventional thermal tension tests. The flow stress of materials can be decreased significantly by increasing the passing current density. For example, at current densities of 18A/mm², 30A/mm², and 35A/mm², the flow stress is reduced by 12.45%, 35.20%, and 48.73%, respectively from room temperature (1100MPa) while the maximum temperature is at 151°C, 383°C, and 736°C, respectively. The resulting microstructures will be examined and reported. Taking advantage of the instant softening behavior under EA-forming, an EA-tube drawing process is proposed to form hollow tubes made of ODS 14YWT steel. A new tubing machine will be designed and fabricated to perform the tube drawing experiments.

Equiatomic solid solution alloys – solid solution metals of several principal elements in roughly equal elemental fractions – are promising candidate materials for advanced nuclear reactor components due to their high radiation resistance. We have studied by means of molecular dynamics simulations the radiation effects in pure Ni and Ni-based fcc alloys with an increasing number of alloying components, up to four in equal elemental fractions. Our combined experimental and simulation results show that several equiatomic metal alloys are more resistant to radiation damage than the corresponding pure elements.Moreover, our analysis of the underlying mechanism of different equiatomic alloys establishes that alloy effects on significant reduction of dislocation mobility is generic and not specific to the number of elements in the system [1]

[1] F. Granberg, K. Nordlund, M. W. Ullah, K. Jin, C. Lu, H. Bei, L. M. Wang, F. Djurabekova, W. J. Weber, , and Y. Zhang, Phys. Rev. Lett. 116, 135504 (2016)

One-dimensional carbon nanotubes (CNTs), which are mechanically strong and flexible, possess better structural properties than their counterpart nanoparticles and other one-dimensional nanowires. We report a nano-dispersion strategy of CNTs in Al matrix that leads to enhance mechanical strength, irradiation resistance with a tenable ductility. This was realized by intragranular dispersion of CNTs from atomic-diffusion driven cold-welding. The uniformly dispersed intragranular CNTs inside metal as 1D fillers (high aspect ratio) create prolific internal interfaces with the metal matrix that act as venues for the radiation defects to recombine (self-heal), reducing void/pore generation and radiation embrittlement at high displacements per atom (DPA). Under helium ion irradiation up to 72 DPA, the 1D carbon nanostructures survive, while sp² bonded graphene transforms to sp³ tetrahedral amorphous carbon. Self-ion (Al) irradiation converts CNTs to a metastable form of Al₄C₃, but still as slender 1D nanorods with prolific internal interfaces that catalyze recombination of radiation defects, reducing radiation hardening and porosity generation. The 1D fillers may also form percolating paths of “nano-chimneys” that outgas the accumulated helium and other fission gases, providing an essential solution to the gas accumulation problem. We further observed nanoscale plasticity and rupturing processes near CNTs by in-situ mechanical tests inside TEM which can contribute to minimize the radiation embrittlement. CNT in the middle of the grain prevents the rupture tip propagation after necking (called Multi-step Rupturing Process (MRP)), which delays the final fracture.
Therefore, our observation of dispersion mechanism of CNT, MPR process and improved radiation resistance deepens scientific understanding and accelerates researches on various CNT-reinforced metal composites.

The modern radiation damage theory of structural materials predicts such experimentally observed phenomena as void and gas bubble super lattice formation, grain boundary and grain size effects etc. In addition it predicts a saturation of void size at very small magnitude in the case when void spatial distribution is random. This phenomenon takes a part in a typical swelling behavior in austenitic steels that normally has the three stages: incubation, transient and steady state. In this talk the effect of void size saturation on swelling accumulation in the dose range up to 100 dpa is studded by use a code RIME (radiation-induced microstructure evolution) based on a mean-field kinetic model, which account for cascade production of one-dimensionally migrating self-interstitial clusters in addition to that of three-dimensionally migrating single vacancies and interstitials. It is shown that the incubation stage is fully determined by the void size saturation. Calculation results obtained for swelling accumulation in austenitic steels at Light Water Reactor conditions are compared with experiment data.

As one of the oldest sciences, alloy development has been focused on traditional alloys with one or two principal elements and minor alloying elements. For nuclear applications, alloy design for enhanced radiation resistance has focused on microstructural or nano-scale features to mitigate displacement damage. In sharp contrast to traditional alloys, recent advances for concentrated solid solution alloys (CSAs) have opened up new frontiers in materials research. In these alloys, a random arrangement of multiple elemental species on a crystalline lattice results in disordered local chemical environments and unique site-to-site lattice distortions. Integrated experimental and modeling studies in a unique set of fcc CSAs with the number of constituent elements varying from 2 to 5 clearly demonstrate links between intrinsic material properties and various dynamic processes. The focus on model fcc alloys eliminates possible confusion or entangled complications from variations of crystal structure, multiphases, or nano-scale features. The systematic variation in the number and concentration of components, as well as the chemical elements, makes this a valuable tactic to de-convolute the effects at electronic and atomic levels. The recent progress suggests a novel concept of how intrinsic chemical disorder can affect defect dynamics at their early stages (low-dose irradiation cases). This has made it possible to build a much-needed knowledge base that may enable reductions in damage accumulation at a later stage (high-dose irradiation cases). The insights into defect dynamics may provide an innovative path forward towards in solving the long-standing challenge in alloy development to achieve much enhanced radiation resistance.

Research supported by EDDE, a DOE-BES Energy Frontier Research Center.

The design of nanoscale microstructures is a promising way to develop radiation-resistant alloys. Since these microstructures are obtained under out-of-equilibrium processing routes they may be unstable under irradiation. An alternative to nanopatterning is to introduce specific solutes which show attractive binding with point defects. These solutes would act as recombination centers, reducing the driving force for void and self-interstitial loop formation. Setting the solute concentrations below solubility limits ensures the stability of the solid solution in most cases. We combine various techniques to study the thermodynamic and kinetic properties of FCC materials (Cu and Al) with added solutes. Starting from the electronic scale, we use density functional theory to calculate binding and migration energies of point-defects and solutes in various local chemical environments. Then we apply the self-consistent mean-field formalism to compute transport coefficients that contain all the kinetic information of the system. Finally we include these transport coefficients in homogeneous rate equations to study steady state point defect concentrations as a function of various parameters. This multi-scale modeling process predicts the efficiency of point defect trapping under different temperature/radiation flux conditions, guiding the design of dedicated experiments.

At ambient temperature, self-induced alpha-decay damage in actinide dioxides and ancient analogues, thorianite and uraninite, (Th,U)O₂, is limited to lattice dilatation of 0.3 to 0.5% even after prolonged exposure to high alpha-decay doses. It is widely accepted that ~1,500 fresh Frenkel pairs are created per alpha-decay event. Most of this damage is attributed to the alpha-recoil atom, traveling a distance of ~20 nm. Here we review two proposed damage recovery mechanisms.
The more important mechanism is radiation annealing - prompt athermal recombination of randomly occurring, closely spaced interstitial-vacancy pairs. Defect ingrowth is modeled as event-by-event production of local damage zones, and simultaneous elimination of close defect pairs. Recombination is governed by a critical interstitial-vacancy annihilation distance. The needed guided defect mobility is controlled by forces that originate from local electrostatic and mechanical distortions. Data pertaining to 18 different actinide dioxide samples yield damage zones having a mean volume of ~3,200 unit cells, and a mean concentration of ~0.1 Frenkel pairs per unit cell at damage saturation. The validity of the model has been confirmed by molecular dynamic simulations on UO₂ at < 5 K. Vacancy and interstitial at closest separation and at some more distant sepatations combine spontaneously in both the uranium and oxygen sublattices.
The second recovery mechanism involves thermal diffusion. Before affected by a subsequent alpha-decay event, a local saturated damage region survives for some time before vanishing into the background of more dispersed damage. This mechanism clearly manifests itself under heating at high temperatures. Evidence at ambient temperature has been provided by leaching of 3 thorianite and uraninite specimens characterized by slow damage rates and lifetime displacements per atom yields of 140 to 450. Annealing is evident by highly enhanced and moderately enhanced initial leaching rates of the short-lived ²²⁸Th and the long-lived ²³⁰Th, respectively, relative to the release rate of the primary very long-lived ²³²Th. These 3 nuclides are expected to reside in fresh, partly annealed and annealed damage, respectively. Similarly, the long-lived ²³⁴U exhibited moderately enhanced leaching rate relative to the leach rate of its primary very long-lived ²³⁸U. It is likely that the ratio of the measured ²²⁸Th/²³²Th and ²³⁴U/²³⁸U activity ratios in excess of congruent isotopic leaching is given by the ratio of the equilibrium fractions of ²²⁸Th and ²³⁴U in residual damage. These fractions depend upon the nuclides decay constants and the damage elimination rate, assumed to decrease exponentially with time. This study yields a mean annealing time of ~15 kyr.
Testing the integrity of freshly-recycled UO₂/PuO₂ reactor fuels under damage created by high loads of minor actinides, and understanding long-term alpha-decay effects in spent UO₂/PuO₂ reactor fuels are of current interest.

Structural materials play an important role in safe, sustainable and economical operation of nuclear power systems. The structural materials for Generation IV concept fission reactors and tokamak-like fusion power plants will be subjected to unprecedented fluxes of high energy neutrons and extreme operating temperature. The development of new high performance materials that can withstand such hostile conditions is, therefore, critical for advanced fission and fusion reactor systems, as well as for high-energy physics. In the search for new materials, a new class of single-phase, concentrated solid-solution alloys (SP-CSAs), might play an important role as structural material for nuclear applications where high radiation tolerance is a primary concern. In this work, we have investigated prolonged irradiation of fcc Ni_0.8Fe_0.2 and Ni_0.8Cr_0.2 alloys using molecular dynamics (MD) simulations to study damage accumulation in these model SP-CSAs, as compared with pure fcc Ni. Recoil energies of 5 and 25 keV are used. The Ni_0.8Cr_0.2 and Ni_0.8Fe_0.2 alloys exhibit higher radiation resistance compared to Ni, and Ni_0.8Cr_0.2 has better radiation performance than Ni_0.8Fe_0.2. In the 25 keV cascade simulations, Ni_0.4Fe_0.4Cr_0.2 is included and compared with the other alloys. For the higher-energy recoil events, the difference in Frenkel pair production between the two binaries is less pronounced; while in the Ni_0.4Fe_0.4Cr_0.2 alloy, defect production is about 1.4 and 2 times lower than the binaries and pure Ni, respectively. While large interstitial clusters are formed in Ni, much smaller interstitial clusters are produced in the alloys with Cr. Due to the low mobility of vacancies on the MD time scales, they are found primarily as single point defects and small clusters in all materials.

Research supported by EDDE, a DOE-BES Energy Frontier Research Center.

A deep understanding of the relationships between chemical composition, structure, and materials properties in the harsh environment of nuclear reactors is of fundamental importance for their safe and sustainable operation. Computational materials science offers unprecedented possibilities to elucidate mechanisms and to predict materials properties as will be reviewed in this talk. Three key factors are driving the progress in computational materials science, namely (i) theoretical concepts such as density functional theory, (ii) computational algorithms and their implementation in software, and (iii) computer hardware. While theory advances steadily on the time-scale of half-centuries, software evolves on the scale of decades, and hardware progresses at a relentless pace of few years. The total impact of computational materials science arises from a combination of these three factors. Thus, the potential value is increasing exponentially. The following examples illustrate current capabilities and challenges. (i) Radiation-induced dimensional changes of Zr alloys can lead to effects such as channel distortion. Simulations reveal that preferential diffusion of self-interstitial atoms lead to the formation of nanoclusters, which deform the surrounding material anisotropically, thus providing an interpretation of experimentally observed radiation-induced growth. Alloying elements such as Nb have a significant influence on these diffusion processes, thereby altering the response of the material to radiation. (ii) With improved corrosion resistance of Zr alloys, hydrogen pick-up and the formation of hydrides with the associated degradation of the mechanical properties are becoming an increasing concern. Computational investigations are helping to unravel the complex mechanisms by allowing detailed simulations of specific scenarios at the water/oxide and oxide/metal interfaces. (iii) Computations provide valuable understanding of the degradation of high-performance alloys. For example, the formation of a Ni₂Cr long-range ordered phase in Ni-Cr alloys causes embrittlement. Computations are revealing the key driving mechanisms, thus setting the stage for mitigation strategies. Atomistic simulations have matured to the point where reliable quantitative predictions of certain materials properties have become possible. This offers a valuable source of materials data such as elastic coefficients, permeability of gases such as hydrogen, segregation energies to grain boundaries, and diffusion rates of alloying elements, impurities, and defects. In addition, simulations are valuable also in the prediction of properties of fluids including density, viscosity, diffusivity and thermal conductivity as needed, for example, in the design and optimization of molten salt reactors. An assessment of the current capabilities and perspectives for future research and development will conclude this contribution.

Ion irradiation is widely used to study the radiation damage of materials for nuclear energy applications, since it costs much less time than neutron irradiation for the same level of displacements per atom (DPA). One-dimensional (1D) SRIM simulation is often performed to estimate the defects distribution by ion irradiation, based on the assumption that target size is much larger than the penetration depth of ion, which usually ranges from a few nanometers to a few microns. However, recent studies push target size down to nano-scale. For example, nanostructured materials, such as nanoporous materials, multilayer nanocomposites and oxide dispersion strengthened (ODS) alloy are promising to be the next generation of structural materials in the nuclear reactors, in that high volume fractions of super sink interfaces facilitate self-healing after radiation damage. For nano-target ion irradiation, full three dimensional (3D) simulations become necessary. Using a recently developed 3D Monte Carlo simulation code for Ion irradiation in Matters (IM3D), we quantify the error of 1D approach in three classic applications of the nano-scale ion implantation. Because 1D approach fails to consider ion exchange at boundaries, a large discrepancy is found between 1D approach and 3D approach. Then by visualization of the 3D defects distribution created by a nano-beam, we discovered a topological evolution for the first time. We conclude that 3D simulation is necessary for nano-scale ion implantation.

The problem of irradiation damage in metals is an important issue in the fields of fusion and fission power generation. High energy neutrons generate large amounts of free vacancies and interstitials in structural materials. The migration of these point defects gives rise to the nucleation and growth of nanometric clusters, such as prismatic dislocation loops and cavities, which modify the mechanical, thermal and electrical properties of the material. Such damage can be recovered in an annealing process, via the climb motion of dislocation loops and evaporation of cavities. These processes are diffusion-driven, meaning that the relevant time scales are much longer than the ones typically achievable with atomistic simulations. In addition, dislocation dynamics (DD) simulations usually employ descriptions of dislocation climb based on mobility laws that disregard non-local effects due to the vacancy distribution, which are likely to be important when investigating configurations with multiple sinks and sources. Here we present a unified continuum model for the annealing of dislocation loops and cavities, which makes use of boundary integral equations to solve the diffusion problem, expanding on recent work by Gu et al. Preliminary results show that non-local effects play an important role when dislocations are closely packed together, or, in general, when a dislocation segment is close to other sources or sinks of vacancies.

Fe-Cr steels are promising candidates for structural materials in advanced fission and future fusion reactors. Magnetism is a key driving force controlling several thermodynamic and kinetic properties of Fe-Cr alloys. Considering the necessity of performing accurate atomistic simulations with large supercells, not accessible with first-principles methods, we have developed a new tight-binding (TB) model for Fe-Cr, where magnetism is treated beyond the usual collinear approximation.

A major advantage of this model consists in a rather simple parameterizing procedure. In particular, no specific property of the binary system is explicitly included in the fitting database. The present model is proved to be accurate and highly transferable for electronic, magnetic and energetic properties in a large variety of structural and chemical environments: surfaces, interfaces, embedded clusters, and the whole compositional range of the binary alloy. The occurrence of non-collinear magnetic configurations caused by magnetic frustrations is successfully predicted.

The present TB approach can also apply to other binary magnetic transition-metal alloys. It is expected to be particularly promissing if the size difference between the alloying elements is rather small and the electronic properties prevail.

The significance of Fe-Cr-Ni system stems from the fact that it forms the basis for austenitic, ferritic and martensitic steels exhibiting diverse magnetic, thermodynamic, mechanical and irradiation properties. Ferritic-martensitic steels like EUROFER and ferritic ODS steels are the most promising candidates for the first wall fusion reactor applications. Despite the fact that Fe-Cr-Ni alloys are one of the most extensively studied ternary alloy systems, the relation between their magnetic phase stability and properties of defects formed under irradiation is not known.
We investigate magnetic properties as well as phase stability of ternary Fe-Cr-Ni alloys at finite temperatures using a combination of Density Functional Theory (DFT) with Cluster Expansion methods. We derive cluster interaction parameters characterizing bcc and fcc crystal structures [1,2]. Enthalpies of formation of ternary alloys predicted by DFT-based Monte Carlo (MC) simulations combined with a correction derived from Magnetic Cluster Expansion at 1600K are in excellent agreement with data derived from experiments performed at 1565K. The relative stability of fcc and bcc phases is assessed by comparing the free energies of alloy formation.
Point defect properties of Fe-Cr-Ni alloys are investigated using representative bulk structures generated using DFT-based MC simulations. The formation energies of vacancies and dumbbell configurations as well as migration energies of vacancies defects in bulk structures are studied as functions of alloy composition the local environment of a defect.
[1] J. S. Wróbel, D. Nguyen-Manh, M. Yu. Lavrentiev, M. Muzyk, S. L. Dudarev, Phys. Rev. B 91, 024108 (2015).
[2] M. Yu. Lavrentiev, J. S. Wróbel, D. Nguyen-Manh, S. L. Dudarev, Phys. Chem. Chem. Phys. 16, 16049 (2014).

High-chromium ferritic/martensitic steels are considered for fuel cladding and other core components in commercial GenIV reactors, as well as for the breeding blanket of fusion systems. However, despite exhibiting an optimal resistance to radiation-induced swelling, the structural integrity, service lifetime and mechanical properties of these materials are threatened, among others, by low-temperature hardening and subsequent embrittlement caused by neutron irradiation damage. To mitigate this problem it is important to understand which radiation-induced nanofeatures are responsible for it, by exposing these materials to irradiation and performing suitable characterization.
The limited availability of suitable neutron irradiation facilities, as well as costs and time expenditure of neutron irradiation experiments, have given rise to a continuously increasing interest in alternative irradiation sources, e.g. ion irradiation. Yet by using a surrogate for neutron irradiation a transferability issue is introduced, mainly due to the displacement rate, which is several orders of magnitude larger in case of ion irradiation, but also due to differences in the damage produced: for example, gradients caused by the self-ion limited penetration which can induce different defect concentrations in the irradiated materials, as well as the injection of extra interstitials.
As part of a multiscale modeling approach, in this work we use an object kinetic Monte Carlo model initially designed for neutron irradiation to study the response of ferritic/martensitic Fe-Cr-C model alloys to a range of conditions. Irradiation temperatures from values relevant for the water-cooled design of the fusion DEMO (i.e. ~250°C) to temperatures of the order of ~400°C, closer to those envisaged for GenIV applications, are explored. The effect of dose-rate and the differences to be expected depending on whether ion- or neutron-irradiation is applied are also investigated. For the simulation of the microstructure evolution under self-ion irradiation, in comparison with neutrons, the irradiation damage profile and the distribution of implanted ions as a function of depth are taken into account, as well as the effect of the ion beam scanning the sample, by comparing continuous versus pulsed irradiation.
Specific reference experiments performed under both ion and neutron irradiation were reproduced and the buildup of the defect populations, in terms of number density and mean size of defect clusters in the model, provided keys to interpret the experimental results.

After a long period without investment, the UK Government is now directly
supporting nuclear materials research. The initiatives including setting up
a National Nuclear User Facility, a Nuclear Fuels Centre of Excellence and
the Sir Henry Royce Institute for advanced materials. The partners involve
UK academic institutions and National Laboratories. Further, in parallel
with this, there has been a real resurgence of interest in the UK in
materials research for nuclear power applications which has stimulated
increased R&D activity in Industry and academia. This has also derived
impetus from significant advances in microstructural and modelling
capability, particularly in academia. The rationale behind these
investments, the recent advances in UK nuclear materials research, and the
vision for future nuclear materials research in the UK will be presented.

To develop fuels with enhanced accident tolerance, the US Department of Energy (DOE) is partnering with fuel vendors to study alternatives to the current UO₂ – zirconium alloy system. The proposed alternative should better tolerate loss of cooling in the core (similar to the Fukushima accident) for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General Electric is proposing to replace zirconium based alloy cladding in current commercial power reactors with a FeCrAl cladding such as APMT. For the last three years extensive testing has been conducted to determine the suitability of the FeCrAl concept under normal operation conditions and under accident conditions. Results show that APMT performs better than zirconium alloys under normal operation conditions and it is several orders of magnitude more resistant to degradation by steam under accident conditions, generating less heat and lower amount of combustible hydrogen. The use of FeCrAl cladding is a direct path to improve the safety of commercial light water reactors.

Nuclear power is one of the largest sources of low-carbon electricity and is considered worldwide as one of the options for mitigating climate change with renewable energy sources. Nuclear fuels play a key role for the objectives of resources optimization, waste minimization, safety and efficiency of nuclear energy and one challenge for the development of future nuclear reactors is to improve significantly the effectiveness of the design and selection of innovative fuels.
Fuel behavior under irradiation is extremely complex due to the combined effect of radiation and temperature. In particular, actinide fission produces large quantities of fission gases which have a significant influence on the structural and mechanical properties of nuclear fuels, as well as chemically active fission products, which induce very complex compositions and interact with other core materials.
An enhanced prediction capability of the fuel behavior in normal operation and accidental conditions is dependent upon improving our detailed understanding of the physical phenomena involved and the induced property changes and upon building more physically based models. Atomic scale methods are efficient tools allowing one to elucidate the elementary mechanisms governing the behavior of materials and their application to UO₂ in recent years has shown that they enable one, combined with separate effect experiments, to get further insight on the behavior of nuclear fuels under irradiation [1-6].
We will show the application of atomic scale simulations to mixed oxide fuels and especially U-Pu oxide fuel or MOX, which will be used in the first European Generation IV prototypes, in particular the French reactor ASTRID. Among the various properties of interest, we will focus on defect and fission gas behavior in MOX and thermodynamic properties. The results will be compared to experimental data from the literature, obtained by X-ray absorption spectroscopy and X-ray diffraction, and to the results obtained in pure UO₂.

References:
[1] B. Dorado et al., “First-principles and experimental study of oxygen diffusion in uranium dioxide”, Phys. Rev. B 83, 035126 (2011)
[2] P. Garcia et al., “Nucleation and growth of intragranular defect and insoluble atom clusters in nuclear oxide fuels”, Nucl. Instrum. Meth. Res. B 277,98 (2012)
[3] J. Wiktor et al., “Coupled experimental and DFT+U investigation of positron lifetimes in UO₂”, Phys. Rev. B 90, 184101 (2014)
[4] P.M. Martin et al., “Behavior of fission gases in nuclear fuel: XAS characterization of Kr in UO₂”, J. Nucl. Mater. 466, 379 (2015)
[5] R. Bès et al., “Experimental evidence of Xe incorporation in Schottky defects in UO₂”, Appl. Phys. Lett., 106, 114102 (2015)
[6] M. Bertolus et al., “Linking scales: atomic and mesoscopic scale modelling of the transport properties of uranium dioxide under irradiation”, J. Nucl. Mater. 462, 475 (2015)

Use of uranium–silicide (U-Si) in place of uranium dioxide (UO₂) is one of the promising concepts being proposed to increase the accident tolerance of nuclear fuels. This is due to a higher thermal conductivity than UO₂ that results in lower centerline temperatures. U-Si also has a higher fissile density, which may enable some new cladding concepts that would otherwise require increased enrichment limits to compensate for their neutronic penalty. However, many critical material properties for U-Si have not been determined experimentally. It is anticipated that modeling and simulation may deliver guidance on the importance of various properties and help prioritize experimental work. In this talk, we will present our recent progress on multiscale material model development for accident tolerant U-Si fuels, spanning density functional theory calculations, molecular dynamics potential development, phase field simulation and engineering scale modeling. Our ultimate goal is to develop knowledge-based models for use at the engineering scale with a minimum of empirical parameters.

As the nations of the world seek to reduce emissions of greenhouse gases and other airborne pollutants, the reliance on carbon intensive fuels will have to be decreased dramatically. The use of nuclear energy as a reliable and scalable option to combat climate change and pollution, while increasing a nation’s energy security, is possible. France provides proof that nuclear energy can displace fossil energy at large scale, where they successfully transitioned from coal to nuclear in two decades. TerraPower has been developing a sodium-cooled fast reactor called the Traveling Wave Reactor (TWR) that aims to greatly improve the perceived shortcomings of nuclear energy by enhancing safety, and providing secure, economic, carbon-free electricity. The TWR will be capable of utilizing either depleted or natural uranium as its main source of fuel. This can be achieved without reprocessing, while reducing the volume of used fuel at the end-of-life. In order to achieve this, the fuel will need to achieve high peak burnups (on the order 30% FIMA), and the materials will need to survive high neutron fluences (on the order of 1.1 x 10²⁴ per cm², or ~550 dpa). These burnups and fluences are much higher than have been achieved historically, and will require test and qualification programs to demonstrate fuel and material performance. TerraPower’s development and test programs are working towards this goal, where both heavy ion and neutron irradiations are being performed, including irradiation of metal fuel. This presentation will describe the program and the results achieved to date.

U_xZrH_y fuels provide a number of design and performance advantageous over conventional UO₂. These advantages include self-moderation (allowing for a more compact core design), enhanced thermal conductivity, and a large negative temperature reactivity coefficient. Both unirradiated and irradiated specimens of a U-Zr-Hydride fuel bonded to Zircaloy cladding with a Pb-Bi liquid metal (LM) were examined with SEM-FIB. The FIB 3D ‘slice and view’ prevented oxidation of the U metal and provided a unique insight into the materials changes following irradiation. Examination revealed significant phase separations in the LM and corrosion/erosion of the cladding and to a lesser extent fuel. In the unirradiated specimen, Pb and Bi appeared to be well structured within the inner LM, Zr appeared to be confined to its expected locations within the cladding and the fuel matrix, and uranium was fairly evenly dispersed within the Zr matrix. The structures in the irradiated specimens were still largely unchanged, but that the inner LM was restructured, the Zr showed some evidence of corroding/eroding away, and there were indications of the presence of Zr in the inner LM. Uranium was also observed to be moving as a bulk phase away from the fuel toward the cladding, smaller amounts of uranium appear dispersed within the inner LM, the fuel surface was fractured and penetrated by inner LM, and the cladding was highly corroded/eroded.

In water-cooled reactors zirconium alloys have been the material of choice for fuel assemblies due to a combination of low neutron cross-section, excellent corrosion performance and good mechanical properties. However, fuel cladding performance, or our ability to predict its performance, remains the limiting factor in an effort to push for increased fuel burnup, i.e. the energy extracted from a fuel assembly before it is removed from the core.

Important in-rector properties of fuel cladding are the resistance to corrosion and hydrogen pick up as well as dimensionally stability of the material. These materials performance aspects are particularly affected by irradiation damage and the associated micro/nano structure evolution, which need to be characterised in great detail.

I will present results of detailed studies using a multiscale characterisation approach by employing diffraction and novel electron microscopy techniques. In order to develop a better understanding of the evolution of damage and micro-segregation, proton as well as neutron irradiated zirconium alloys have been investigated using STEM-based ultrahigh resolution EDX spectrum imaging and synchrotron x-ray diffraction. These investigations have been carried out on two types of zirconium alloys, Zircaloys (no Nb) and Zr-Nb type alloys. These studies have provided a detailed picture and have enabled quantitative analysis of the evolution of and dislocation loops as a function of dpa. In addition, ultrahigh resolution EDX mapping provides new insight in the possible role of micro-segregation and the formation of nano-precipitates or nano-clusters on dislocation loop formation. These new observations are interpreted in the view of dimensional instabilities observed for zirconium alloys, particularly growth, as well as the potential impact on corrosion and hydrogen pick-up

The use of ion beam irradiations as a meaningful emulation tool for high-dose neutron irradiations is rapidly being qualified in the nuclear materials community. However, current schemes of materials examination are largely unchanged from the standard processes developed for neutron exposures. Namely, samples are characterized, irradiated, and then undergo post-irradiation examination (PIE). Given the flexible, low-activation environment present in many ion beam facilities, the opportunity exists for the development of continuous, on-line characterization techniques to monitor material property changes in-situ, as irradiation is taking place. The use of transient grating spectroscopy (TGS), a photoacoustic technique sensitive to several thermo-mechanical properties of interest in irradiated materials, is being developed for this purpose. Initial ex-situ characterization of self-ion irradiated pure single crystal and polycrystalline copper samples showed distinct changes in measured TGS signals, related surface morphology changes, and a reduction in elastic modulus and thermal diffusivity. These initial studies motivate the development of an in-situ TGS ion beamline experiment, the conceptual design of which has been completed. Such a facility would allow for not only vastly increased dose resolution of material properties affected by irradiation, but also for targeted PIE at transition points in irradiated material response, such as at the onset of breakaway void swelling. This technique therefore has applications as a tool to pinpoint the onset of radiation void swelling, allowing for the far more rapid qualification and study of families of alloys designed for extreme radiation resistance.

Ion-irradiation is an important method for damage simulation with short term experiment, but a disadvantage is inhomogeneous damage profile as a function of depth. In this study the multi-layer nano-indentation method was applied for evaluating the dose dependence of radiation-hardening from a piece of irradiated specimen. Austenitic and F/M steels were irradiated by 10.5 MeV Fe³⁺ to 20 dpa (in average) at 300 C.
The range of plastically deformed region due to the nano-indentation was confirmed by Nix-Gao plots and transmission electron microscopy. The multi-layer nano-indentation method was based on the following assumptions; (1) the irradiated region can be divided into sub-layers having their own local hardness, (2) the local hardness can be depended on the fraction of each sub-layer, and (3) the deformation zone is a hemisphere shape.
Eventually, through the FIB sectioning followed by nano-indentation, local hardness in each sub-layer was experimentally evaluated, and then we got a dose dependence of radiation-hardening RH up to 60 dpa with reasonable relation of H_V = A + B (dose) ⁿ. Furthermore the correlation between ion- and neutron irradiation data will be discussed in this study.

There is a strong need to develop advanced instrumentation for measurement of thermal properties of nuclear materials in extreme environments for the analysis of the operation limits of nuclear fuels, prediction of possible nuclear reactor accidents and for development of next generation safe and efficient nuclear fuels. Thermal arrest of nuclear fuels suffers from fuel-furnace interaction along with volatilization and contamination during melt testing. To minimize these problems an ample progress has been made in developing experimental techniques enabling investigating the response of nuclear fuels to millisecond laser pulse heating. In this work we explore the response of refractory materials to nanosecond pulse laser surface heating, combined with in-situ laser vibrometry based on two-wave mixing photorefractive interferometry technique.
We demonstrate automated nanosecond pulsed laser heating of metals across melting, but below ablation regime. We have measured laser pulse-induced melting thresholds for different refractory metals by detecting laser-generated ultrasound. The experiments were conducted with motorized stepper motor to control the incident laser pulse energies and simultaneously measure both the varying laser pulse energy and epicentral acoustic waveforms. Finite element modeling based on coupled elastodynamic and thermal conduction governing equations were found to be in good agreement with experimental results. The onset of melting was attributed to the formation of the molten pool leading to preferential generation of shear waves from acoustic sources surrounding the molten mass resulting in the delay of shear wave arrival times. Observed trend in the increased melting thresholds for given refractory metals is in agreement with the increase of their corresponding melting point temperatures. It was also observed that the relative delay in shear wave arrival times above melting threshold increases in metals with higher melting points.
Demonstrated pulsed laser technique has proven to be a suitable and promising to study high temperature phase transitions. A nanosecond pulse laser melting technique using in-situ laser vibrometry can be installed in a hot cell where time- and spatially-resolved examination of solid-to-liquid phase transition in fresh and irradiated fuel samples can be examined. In-situ laser vibrometry based on two-wave mixing photorefractive interferometry determining phase transitions in refractory metals provides an independent, reproducible test to determine the onset of melting and can investigate the high temperature properties of nuclear materials. Latest measurement results on pulsed laser melting thresholds in various refractory materials will be discussed in the light of acoustic and thermal phenomena.
Acknowledgement
This research was funded by Nazarbayev University and target program № 0115РК03029 "NU-Berkeley strategic initiative in critical state of matter, advanced materials and energy sources" from the Ministry of Education and Science of the Republic of Kazakhstan.

Advanced reactors will require development of new alloys that can withstand extreme environments, including high temperatures and neutron displacement damage, and must also manage large amounts of He generated in neutron-alpha reactions. These challenging service conditions lead to microstructure changes on the nano-scale, such as precipitation or He bubble/void formation, that are often at the resolution limit of characterization techniques. Atom probe tomography (APT) provides high resolution compositional measurements that track changes in solute locations, such as precipitation and segregation, in what are intrinsically 3D tomographic reconstructions. TEM provides other information, such as crystal structures, and probes larger sampling volumes, but often only in 2D projections. Here we report on insight gained using a correlative combination of APT and TEM tomography on the same sample volumes. Data is presented for tempered martensitic steels containing He bubbles and voids and nanostructured ferritic alloys containing nano Y-Ti-O oxide precipitates, and precipitate associated He bubbles. The correlative combination of these methods provides unique information and understanding of nanostructures and their associations. In addition, the correlated datasets highlight artifacts associated with the individual techniques.

While many graphite-moderated nuclear reactors in the world have seen their production times extended well above initial predictions, there is a crucial need to better understand the effect of prolonged irradiation on the structure of graphite parts as damages might both affect the mechanical integrity of the cores or perturb their thermal management.
In this work we have used electrons as a surrogate for neutron irradiation and have combined in-situ high-resolution transmission electron microscopy (HRTEM) analysis of electron irradiation in graphite with an image-based atomistic reconstruction procedure [1], which allows us to propose a dynamic time series of atomistic models, up to a total dose equivalent to approximately four months in a reactor core. We show that even at such high doses, the material retains highly anisotropic grains maintaining anisotropic properties such as stiffness and thermal conductivity. The proposed models, after property homogenization, account for the experimentally observed increase in Young's modulus and decrease in thermal conductivity with increasing dose. Further validation of the models is provided via a comparison of simulated and experimental data from irradiated material such as: HRTEM images, carbon K-edge electron energy loss spectra, dose rate and stored energies [2].
[1] Leyssale et al., App. Phys. Lett. 95, 231912 (2009); Farbos et al. Carbon 90, 472 (2014).
[2] Farbos et al. Under review

High-entropy alloys (HEAs) have recently been developed as nontraditional alloy systems. They are composed of multiple elements at or near equiatomic ratios that form random solid solutions on simple underlying lattices. In recent years HEAs have attracted significant attention due to their high strength, ductility and possible high radiation resistance. As a prototype system of HEAs we choose Ni-based binary and ternary alloys.
Energy transfer between lattice atoms and electrons is an important channel of energy dissipation during displacement cascade evolution in irradiated materials. On the assumption of small atomic displacements, the intensity of this transfer is controlled by the strength of electron– phonon (el–ph) coupling. The el–ph coupling in concentrated Ni-based alloys was calculated using electronic structure results obtained within the coherent potential approximation. It was found that Ni_0.5Fe_0.5, Ni_0.5Co_0.5, Ni_0.5Pd_0.5, Ni_0.33Co_0.33Cr_0.33 and Ni_0.4Fe_0.4Cr_0.2 are ordered ferromagnetically, whereas Ni_0.5Cr_0.5 is nonmagnetic. The el–ph coupling values for all alloys are approximately 50% smaller in the magnetic state than for the same alloy in a nonmagnetic state. As the temperature increases, the calculated coupling decreases. The rate of decrease is controlled by the shape of the density of states above the Fermi level. Introducing a two-temperature model based on these parameters in 10 keV molecular dynamics cascade simulation increases defect production by 10–20% in the alloys under consideration.
This work was supported as part of the Energy Dissipation to Defect Evolution (EDDE), an Energy Frontier Research Center funded by the U.S. Department of Energy, Office of Science, Basic Energy Sciences. LKB acknowledges additional support from a fellowship awarded by the Fonds Québécois de recherche Nature et Technologies.

Tungsten-base materials are currently the most promising candidates for divertor armor components in future fusion reactors. During service they will be exposed to high temperatures, intense neutron flux and ion bombardment. Here we use helium ion implantation to study the interaction of injected helium with displacement damage and the effect the resulting defects have on the mechanical and physical properties of tungsten.

Using X-ray micro-diffraction and laser-induced transient grating measurements, we observe both lattice swelling and changes in the elastic modulus after ion implantation. Surprisingly, a fraction of a percent lattice expansion, driven by the accumulation of helium and implantation defects, causes an order of magnitude larger reduction of the elastic modulus. These observations are interpreted using a combined elasticity and density functional theory based model. We find that our experimental observations are consistent with a damage microstructure dominated by Frenkel defects with helium-filled vacancies. Measurements on helium-implanted single crystals further confirm the slight increase in elastic anisotropy predicted by our calculations.

Laser-induced transient grating measurements can also determine the thermal transport properties of the implanted layer. We find that even a modest concentration of injected helium leads to substantial reductions in thermal diffusivity. Using a kinetic theory model this effect can be explained in terms of the implantation-induced damage microstructure. Importantly the changes in thermal transport properties are not a trivial function of the implanted ion dose.

X-ray micro-diffraction measurements of lattice strains in samples heat-treated after ion-implantation show significant evolution of the damage microstructure. Indeed they suggest that at elevated temperatures defects migrate deeper into the material bulk. These multifaceted observations combined with multi-scale modeling allow us to begin to establish a joined-up picture of the complex effects helium-implantation-induced damage has on the structure and properties of tungsten.

The displacements per atom (DPA) is the current concept to quantify radiation damage in materials. The DPA of a material is defined as a calculation, not a measurement, integrating the energy-dependent fluence of energetic particles multiplied by the damage displacement cross section, and is a measurement of the number of times each atom is displaced from its original lattice site. It therefore is a measure of ballistic radiation damage, not the actual time-evolved residual damage that determines irradiation-affected material properties. In addition, no method exists to measure the DPA of a material irradiated in an unknown condition, as it is a fundamentally unmeasurable quantity. Methods such as transmission electron microscopy (TEM), atom probe tomography (APT), and mechanical testing can reveal some of radiation’s effects on materials, but cannot directly yield the defect populations which fully define irradiated material properties.

In order to gain a more fundamental understanding of radiation damage, a measurable unit of radiation damage must be developed which accounts for other irradiation parameters, such as dose rate and irradiation temperature. We propose the Wigner energy spectral fingerprint, a calorimetric measure of defect energy release during annealing at multiple temperatures, to quantify material damage. This directly measures the potential energy stored in defects in materials, by radiation and/or by other means. A combination of molecular dynamics (MD) simulations and calorimetric experiments shows that the Wigner energy approach quantifies material damage, and correctly identifies defect types by their stored energy per atom.

We present the results of neutron irradiation experiments and MD simulations of radiation damage in six different materials (Al, Cu, W, Zr, Fe, 304SS), spanning a wide range of crystal structures and melting points. MD simulations show that the total stored Wigner energies follow a scaling law normalized by the cohesive energy of the material, while defect types differ with varying PKA energy and material. Total stored energy is found to increase with PKA energy, following a power law with an exponent of about 0.85. Finally, we show that histograms of stored energy per atom are useful to automatically determine defect populations in irradiated materials. We also show the results of the experimental irradiation of these six materials by fast neutrons in the MIT reactor, along with differential scanning calorimetry (DSC) measurements of stored energy spectra. Heating rates ranging from 2°C/min to 2,000,000°C/min reveal different aspects about the Wigner energy spectral fingerprint of each material.

Applications of using the Wigner energy spectral fingerprint include determination of unknown dose levels of irradiated materials, providing fundamental understanding about the nature of radiation damage, and verifying historical production of enriched uranium for non-proliferation and treaty verification.

Grain growth in irradiated elemental nanocrystalline (nc) metals follows a power law in which the grain size depends on the dose. The prevailing models assume that the grain growth is due to a thermal spike of ~10 ps duration that follows the collision and relaxation phases. In these models, the driving force for growth is due to the chemical potential difference resulting from grain boundary energy divided by grain boundary curvature, per the Gibbs-Thompson expression. The main features of the models are that the power law exponent, n, is quantitatively predicted (n~3) and that grain growth continues as dose increases. In thermally stabilized nc NiW, it is believed that the grain boundary energy is reduced, thus enhancing its resistance to grain growth at elevated temperature. According to the thermal spike model, irradiated NiW grains should then grow more slowly than pure Ni grains. However, our experimental results are in striking contrast with this expectation. Using Ni self-ion irradation, up to 10 displacements per atom (dpa), NiW nc grains grow much faster than Ni nc grains. The grain size is linear with irradiation and grows from 6 to 50 nm (n~1). However, beyond 10 dpa and up to 100 dpa, no further grain growth is observed (n>10 !). To investigate the source of the abrupt halt in grain growth, we are conducting TEM-based precession diffraction, by which we index the phase and orientation of each grain to gain quantitative information on the nc-metal texture evolution and emergence of the dominant grain boundary types. Initial results indicate a high fraction of low angle grain boundaries, which are likely to resist grain growth. This work will lead to a better understanding of irradiation-induced grain growth mechanisms in thermally stabilized nc metals, as well as conditions required to suppress grain growth.

The development of candidate ferritic-martensitic alloys and ODS variants of these alloys for advanced reactor service is hampered by the lack of neutron irradiation facilities capable of reaching the high doses required to assess their microstructural and dimensional stability under prolonged irradiation. One approach to overcome this problem is to use charged particle irradiation at accelerated displacement rates, with self-ions best suited for this purpose. This paper presents the results of a series of comparative irradiations focusing on void swelling, phase stability and the stability and action of various types of dispersoids. Two major accelerator facilities are involved, with some alloys irradiated in both facilities. One facility is in Kharkov, Ukraine and uses 1.8 MeV Cr ions. The second facility is in College Station, Texas and uses 3.5 MeV Fe ions.

The investigated ferritic-martensitic alloys are several heats of HT9, several heats of EP-450, T91 in several TMT variants, F82H, GA3X, ChS-139, EI-852, EK-181 and several nano-structured variants of EK-181. The ODS alloys are EK181-ODS, EP-450-ODS, MA956, several heats of MA957, 14YWT, PM-2000, FeCrAl-ODS and two groups of unique ODS steels developed at Hokkaido University and Kyoto University. Additional alloys are being added as they become available. Depending on the observed swelling resistance of each alloy, the dpa levels investigated range from 200 to 1000 dpa. While 1000 dpa might seem to be an unrealistically high dose, there is a suppression of void nucleation in the injected interstitial zone that precludes the extraction of swelling data in the higher-dose region of the damage vs. depth curve, producing ~600 dpa as the upper dose limit from a 1000 dpa peak dose.

All investigated alloys eventually exhibit void growth, with ODS steels in general, but not always, resisting the onset of swelling longer than do non-ODS counterparts. Those alloys that exhibit accelerated void growth swell at ~0.2%/dpa, but with vastly different incubation periods. MA 956 is an example of an ODS alloy where dispersoids actually accelerate the onset of swelling.

The results of these studies point to several routes to develop improved swelling resistance. Notably, dispersoid stability appears to be rather variable. One significant result is that the major role of dispersoids for swelling suppression appears to be maintenance of the stability of nano-structured grains with significant denuded zone. Several examples are shown where maintenance of nanostructured grains resists the transition to accelerated void growth to at least 500 dpa.

The limited availability of high flux neutron irradiation sources severely slows development of improved alloys for advanced reactors that require structural materials displaying radiation resistance to very high dpa levels of 300-600 dpa. Therefore the materials community has turned to charged particle irradiation at accelerated dpa rates as a surrogate for neutron irradiation to speed up the development process. This technique has especially focused on void swelling, a non-saturable process that can produce very large volume changes and distortions at high damage levels.
Void swelling can be divided into two major regimes, void nucleation which dominates the transient regime, and post-transient high-rate swelling regime, often referred to as "steady-state" swelling. While the latter is not very sensitive to most environmental and production variables, the former is very sensitive to almost all variables. It is well-known, however, that there are some significant differences, usually referred to as "neutron-atypical" aspects of ion irradiation that complicate the extrapolation of ion-induced swelling to neutron-relevant applications. These aspects not only add additional complexity but sometimes strongly suppress void nucleation.
One of the strongest differences between ion and neutron irradiation is the orders of magnitude difference in damage rate. Not only is the void nucleation process strongly dependent on dpa rate even in the simplest of alloys, there is often a radiation-driven microchemical evolution of the alloy matrix that precedes and defines the duration of the void nucleation regime. The dependence of the microchemical evolution is also sensitive to dpa rate, but not necessarily in the same manner as that of void nucleation, with the consequence that ion-induced swelling is more difficult to extrapolate to neutron-relevant conditions.
One approach to mitigate this problem is to use neutron-irradiated or "preconditioned" material at some moderate damage level and then continue the irradiation with self-ions to larger damage levels. In this case the microstructure and microchemistry of the alloy were already moving in a direction dictated by the neutron environment. Additionally, if void nucleation is largely in progress or complete, then suppression of nucleation by neutron-atypical aspects of subsequent ion irradiation are much less relevant.
In the 1970s and 1980s a series of such preconditioning experiments were initiated to explore the utility of the technique and to define how to select combinations of temperature and dose rate that would yield the best simulation. Many of these experiments were not published or completely analyzed, however, and the communities' perceptions of how to calculate ion-dose profiles were later strongly revised. This paper reviews and reanalyzes these experiments to provide guidance on how best to use this neutron-plus-ion technique for better simulation of void swelling at very high dpa levels.

Reactor core materials in both fast reactors and LWRs granted life extension must withstand irradiation to high doses at high temperature. Ferritic-martensitic (F-M) alloys are attractive candidates for structural components of fast and thermal reactors, and high Cr and high Ni stainless steels are potential replacement alloys for LWR core materials. To reach high damage levels, self-ion irradiation is conducted with simultaneous He injection into both F-M and austenitic alloys, accompanied by reactor irradiations conducted in the FFTF and BOR-60 fast reactor to assess the capability of ion irradiation to emulate the evolution of microstructure and mechanical properties in reactor. The evolution of microstructure at high damage levels and the interrelationships of the various features will be presented, along with comparisons to reactor irradiated microstructures.

It is well known that the BOR-60 fast reactor can accumulate only 20 displacements per atom (dpa) or less per year in iron-base structural alloys. No other fast reactors are currently available. Currently available mixed-spectrum reactors (HFIR, ATR) in the USA cannot accumulate more than 10 dpa per year and usually much less in available test volumes. Total damage levels of 200-600 dpa are required to test the swelling resistance of advanced ferritic-martensitic alloy and their ODS variants. Therefore the international radiation materials community has turned its attention to charged particle simulation techniques at vastly accelerated dpa rates as surrogates for evaluation of new alloy concepts and candidates, especially with respect to void swelling and irradiation creep. However, there are significant differences between the spatial and temporal environments of neutron and charged particle irradiations. The various neutron-atypical aspects of ion bombardment preclude, however, an exact one-to-one ion-neutron comparison for confident near-exact prediction of swelling in a neutron environment. Claims were frequently made that predictive capability was possible, but such claims have not been validated and should be taken with a grain of salt.
In this talk, the most important "neutron-atypical" characteristics or "artifacts" of charged particle irradiation will be discussed. Major atypical factors such as defect imbalance, defect gradients, rastering and displacement rate effects are fairly well known and are being appropriately considered in data analysis. For each neutron-atypical factor, experimental evidence, multi-scale modeling, and solution are discussed in this talk. A call is made for urgent attention on standardization of accelerator testing procedures to maximize the credibility of any neutron-ion correlation.
However, as we optimize the ion irradiation to minimize the impact of some variables such as rastering, we can open the door to new challenges. In this presentation we will focus especially on a recently identified new issue, carbon injection from the accelerator beam lines into the irradiated area. Carbon contamination has not been well studied and is usually assumed to be ignorable. However, in non-rastered irradiations carbon contamination can increase significantly, and carbon is well known to participate strongly in void swelling due to strong defect interactions with carbon atoms or carbides. The issue is serious enough to possibly invalidate high dpa testing data of void swelling in stainless steels. We show that near-surface carbon enrichment results from beam-induced surface contamination, and we provide guidance on specific instrumental innovations to solve the contamination issue.

Solid-state interfaces in metallic composites are preferential sites for helium (He) precipitation. We aim to predict the distribution of He network precipitates at interfaces to improve performance of plasma-facing materials in fusion devices. We use O-lattice theory to determine the initial distribution of He precipitate nucleation sites and phase-field methods to model the growth, interaction, and coalescence of He bubble networks. These methods can be used to design next generation multilayered metallic composites with enhanced resistance to He damage. We demonstrate our computational methodology on a Cu-V interface [1]. We compare our simulations to experimental TEM images of He implanted Cu-V interfaces.


[1] D. V. Yuryev and M. J. Demkowicz, "Computational design of solid-state interfaces using O-lattice theory: An application to mitigating helium-induced damage," Applied Physics Letters, vol. 105, Dec 1 2014.

A key goal of materials research in the realm of nuclear energy is to design new materials that can withstand the high doses and extreme temperatures expected of next-generation nuclear reactors. Thus, there has been significant effort to understand the factors that control radiation damage evolution. Several such factors have been identified, including the use of interfaces, both grain boundaries and heterophase interfaces, as sinks to enhance the annihilation of defects produced under irradiation. However, despite a long-standing realization that interfaces can promote radiation tolerance, the precise atomic-scale mechanisms responsible have been murky at best.

Using a combination of accelerated molecular dynamics and kinetic Monte Carlo simulations, we examine defect interactions with grain boundaries in copper, a model system for more complex materials. We find that both the thermodynamics of defect-interface interactions as well as the kinetics of defects at the interfaces themselves are very sensitive to the atomic structure of the interfaces. Thus, the mobility of both interstitials and vacancies vary significantly from interface to interface. This, in turn, impacts the annihilation rate of defects within the interface. Effectively, each interface has a different defect annihilation rate which dramatically affects its sink efficiency and thus its ability to enhance radiation tolerance. We conclude that defect mobilities are interfaces are a key factor in determining whether a given interface is able to improve the overall response of the material to irradiation.

Understanding the self-healing mechanisms of defects in nanocrystalline materials is of importance for developing the structural materials that can support extended component lifetime under extremely hostile conditions. It has been revealed that grain boundary can play a critical role in the radiation damage process by acting as sinks of various point defects. In the present work, first principles calculations are carried out to investigate the energetic landscape of point defects (i.e. self-interstitials, He-interstitials, and vacancies) induced by the irradiation damage and their migration properties of the self-healing process in the vicinity of grain boundaries (GBs) in copper, focusing on six symmetric tilt grain boundaries that vary in their energies. Our results indicate that the low-energy GBs are generally accompanied by a higher propensity of self-healing behavior due to the inter-granular interstitials and intra-granular vacancies recombine with each other. The recombination process is proved to be regulated by two mechanisms: interstitial emission mechanism and vacancy mediated mechanism. For the low-energy GBs, the former mechanism demonstrates its efficiency in describing the atomic motion, while for the high-energy ones, the latter turns out to be superior. With the aid of these mechanisms, we conclude that the low-energy GBs are comparatively more radiation-resistant than the high-energy counterparts, which may shed light on the rational design of high-performance structural materials based on nanocrystalline alloys.

Tungsten (W) is prime candidate material for a plasma facing components in a fusion reactor. The degradation of the material properties of W due to high heat load and neutron irradiation is predicted during fusion reactor operation. To increase the reliability and lifetime of W, improvement of the mechanical property and the resistance against irradiation and recrystallization embrittlement are needed. W alloys such as potassium (K)-bubble dispersed and/or rhenium (Re) added W were designed based on the results of researches on neutron irradiation and heat treatment effects on microstructure and material properties of pure W and its alloys. Materials examined in this work were supplied by A.L.M.T. Corp., Japan, and was fabricated by powder metallurgy and hot rolling. The microstructure observation and the measurement of the mechanical and thermal properties were performed. The experimental results showed that the finer grain structure and higher tensile strength of W alloys than those of pure W, and the effectiveness of combination of second-phase dispersion and addition of alloying element was found. One of the drawback by addition of alloying element was decrease in thermal property, although this was not occurred by second-phase dispersion. To clarify the trade-off relation between improved and worsened properties, the structural analysis by using finite element method was performed to evaluate stress and strain responses of pure W and its alloys under fusion reactor relevant heat load conditions, and details of this analysis will be presented. To evaluate the long-term reliability of pure W and its alloys, the effects of thermal history and heavy-ion irradiation on microstructural development and mechanical properties such as tensile property and hardness were investigated. The details of the experimental results and the perspective of further high performance W alloys development will be presented.

Tungsten is one of the primary plasma-facing materials used in fusion technology. Exposure to a plasma of hydrogen isotopes leads to surface and sub-surface modification of tungsten-based materials, such as formation of macroscopic sub-surface cavities resulting in surface blistering. Plasma-induced modification is locally non-uniform – laterally and in depth, with a characteristic length-scale on the order of micrometers.

Several techniques are able to probe the relevant mechanical and thermal properties of tungsten locally at corresponding length scales. Continuous stiffness measurement nanoindentation allows depth-resolved probing of mechanical properties of individual grains in polycrystalline materials. Laser-induced transient grating measurements allow determination of local near-surface (within ~1 µm from the surface) thermal diffusivity. This contribution reports new results applying these techniques, combined with SEM surface imaging, EBSD crystallographic analysis and FIB cross-sectioning, to tungsten exposed to deuterium plasmas under the divertor-relevant conditions – ion flux density ~10²⁴ m^-2s^-1, ion energy ~50 eV. We will address both the impact of intrinsic properties on the effects of exposure, and the effect of exposure on material properties.

We find that plasma exposure increases material hardness by up to ~1 GPa. The measured hardening decreases with depth, following the diffusional profile of deuterium concentration. Hardness is independent of crystallographic orientation, both in as-received and exposed material, but depends on the degree of internal misorientation within individual grains, being up to ~20% higher in deformed grains compared to recrystallized ones. Lateral dependence of blistering is found to depend on the polishing state of the surface. Mechanically ground samples exhibit a strong crystallographic dependence of blistering (with grains with <111> surface normal orientation the most susceptible to blistering and those with <001> blister-free), while surfaces of electrochemically polished samples form blisters regardless of grain orientation. On the other hand, blister size and areal density depend on internal misorientation, with larger densities of smaller blisters present in deformed grains. Plasma exposure also degrades thermal transport properties, reducing thermal diffusivity in near-surface region by up to ~15%.

The components and materials facing the plasma in a fusion reactor are subject to extreme and cyclic particle, power and neutron loads. The unique combination of loading conditions requires new “advanced” materials and components in order to realize fusion as an energy source for the future. High heat fluxes and hence surface temperatures establish strong thermal and stress gradients in actively cooled components. The application of different materials as armor towards the plasma, heat sink and structural materials or bonding layers create potential failure zones in these components, especially under cyclic loads. As a very important aspect, hydrogen isotope diffusion and permeation as well as erosion have to be considered in the material selection. The design of a fusion reactor requires a large degree of inherent safety in case of a potential loss-of-coolant event with air ingress.

To accommodate these combined material and system requirements, new material concepts are currently developed and neutron effects in fusion materials are studied by using ion and neutron beams. Damage induced by radiation deteriorates cycle stability and thermo-mechanical properties of the materials. Only in irradiation studies a credible lifetime and operational temperature window of plasma-facing materials can be assessed, as the deterioration is strongly material and temperature dependent.

This talk provides an insight in specific material requirements for first wall materials in future fusion power plants. As examples of current research, the development of tungsten fiber-reinforced tungsten composites, new self-passivating tungsten alloys and effects of nuclear damage (induced by fast particles) on material properties will be discussed. A new “advanced” material concept for a fusion reactor is envisaged, in particular the combination of oxidation-resistant tungsten alloys – which are generally even more brittle than pure tungsten – with a composite concept to establish pseudo-ductile behavior and controlled crack growth. Manufacturing routes of tungsten alloys and composites include powder metallurgy, as the classical route to refractory metals, as well as chemical vapor deposition and infiltration techniques. Further, experiments with >15 MeV proton beams to simulate irradiation damage by 14 MeV fusion neutrons are presented. The advantages and challenges of this technique in comparison to the application of low-energy fission neutrons will be outlined and discussed.

Concerning nuclear fusion systems, namely ITER (International Thermonuclear Experimental Reactor) and DEMO (DEMostrator Reactor), which take the tritium deuterium reaction as a reference, tritium breeding from Pb-16Li eutectic is one of the focus point of technological R&D activities worldwide. The permeation of tritium through the foreseen steels of eutectic Pb-16Li Breeder Blanket (BB) concepts is a challenge for the tritium balance in the reactor. Once tritium is produced, it must be appropriately extracted from the Pb-16Li eutectic, precluding losses by permeation through the steels towards the Helium or Water cooling systems. In order to avoid such losses, an adequate permeation barrier is required. We report on the barrier performance of advanced Al₂O₃ ceramic coatings, which are suitable for this task owing to their chemical inertia, high density and amorphous character.
In particular, hydrogen gas is used in order to simulate tritium. Tests are performed with a fixed barrier thickness (5 µm) and under a fixed hydrogen partial pressure (100 mBar). The temperature of the tests is varied within the range 523-923 K. Hydrogen permeation through coated disc specimens is detected using a quadrupole mass spectrometer. Our results show that the Al₂O₃ coatings tested are a promising tritium permeation barrier, with permeation reduction factors (PRF) approaching 10⁵. Last, but not the least, we show that the coatings perform well as anti-corrosion barriers in harsh environments relevant for BB concepts, such as liquid lead.

In future nuclear fusion devices, plasma-facing components (PFCs) will be subjected to high thermal loads and high-fluxes of helium (He) and hydrogen (H) isotopes. Currently, tungsten (W) is considered to be the leading candidate PFC due to its high melting temperature, high thermal conductivity, and low sputtering/erosion rates. However, recent studies have shown that W undergoes significant surface morphology evolutions (blisters, bubbles, and tendril-like nanostructures (fuzz)) when irradiated with high fluxes of low-energy He⁺ ions under simultaneous sample heating at elevated temperatures [1]. These nanostructures, in turn, could lead to an additional source of plasma impurities and negatively impact reactor performance. The motivation of this work is then to explore He⁺ ion irradiation effects in other candidate PFC surfaces. Namely, our preliminary results for He⁺ ion irradiation of tantalum (Ta) have shown that Ta may exhibit a higher fluence tolerance for nanostructure formation [2]. Mirror-polished Ta samples were irradiated with 100 eV He⁺ ions at a flux of 1.2 × 10²¹ ions m^-2 s^-1 to total fluences in the range 4.3 × 10²⁴ - 3.5 × 10²⁵ ions m^-2 at constant temperatures in the range 773-1223 K. Resulting surface morphology changes were monitored and characterized with a combination of scanning electron microscopy (SEM), focused ion beam (FIB) SEM, X-ray photoelectron spectroscopy (XPS), X-ray diffraction (XRD), and specular optical reflectivity. Overall, it was found that Ta has similar structure formation temperature regimes to W and forms similar blister and tendril-like surface structures. However, Ta demonstrates a higher fluence tolerance to fuzz-like nanostructure formation. While Ta is often discredited as a PFC due to its relatively low thermal conductivity compared to W, this tolerance to high aspect ratio surface structure formation suggests that the thermal conductivity of Ta is less likely to degrade with time in an extreme reactor environment, therefore reclaiming its relevance as a PFC.

[1] S. Kajita, N. Yoshida, N. Ohno, Y. Tsuji, New. J. Phys. 17 (2015) 043038
[2] T.J. Novakowski, J.K. Tripathi, A. Hassanein, J. Nucl. Mater. 467 (2015) 244-250.

Thanks to its high melting point and thermal conductivity, tungsten (W) is among the most promising candidates as plasma facing material (PFM) in fusion reactors (e.g. ITER, DEMO), where the resistance to extreme thermal loads during steady and transient operating conditions is a key requirement. Because of their peculiar properties, which can significantly differ from the bulk ones, W coatings are also widely exploited in Tokamaks as protective layers for plasma facing components (e.g. WEST). In addition, amorphous-like W coatings (a-W), with crystallites dimension < 1 nm and mass density about 11 g cm^-3, have been used at lab-scale as proxy of redeposited W in the framework of plasma-wall interaction, mimicking bulk W modification during different tokamak scenarios [1].

In this work, we deposit a-W coatings by Pulsed Laser Deposition (PLD). Thanks to its high versatility in tuning various deposition parameters, PLD allows us to finely control the growth process of the coatings, tailoring their structure down to the nanoscale. The as-deposited a-W samples are then processed by thermal annealing treatments at temperatures up to 1200 K. The recrystallization behavior and its effects on the mechanical properties are thus investigated. The mechanical properties are measured by Brillouin spectroscopy, while X-ray diffraction analysis and Scanning Electron Microscopy monitor the structure and morphology evolution. We find that the recrystallization process for a-W starts at around 900 K, which is well below the bulk W recrystallization temperature (1700 K), where an increase by 60% of material stiffness with a corresponding loss of ductility by about 30% is observed [2].

As proposed in [3], the extreme thermal loads proper of nanoseconds laser pulses can be exploited to simulate the thermal effects of plasma transient events (i.e. ELMs) on W. In addition, the thermal fatigue effects and failure mechanisms (e.g. cracking, delamination) on a-W coatings can be investigated by fast thermal annealing induced by nanoseconds laser irradiation. In view of an experimental campaign of this type, we developed a numerical tool aimed at supporting the design of the experiments and at interpreting the measured data. The temperature and stress fields within the multilayer structure, and their evolution during irradiation, are thus derived by a 2D axial-symmetric numerical code based on the finite difference method. Particular attention is paid to the modeling of the interface between adjacent layers. Taking into account the ductile-to-brittle transition temperature of W, different conditions of cracks formation can be numerically deduced. The total coating/substrate interface stress is also derived so that the delamination process can be investigated.

[1] M.H.J. ‘t Hoen et al., Journal of Nuclear Materials 463, 989 (2015)
[2] E. Besozzi et al., Materials and Design 106 14–21 (2016)
[3] N. Farid et al., Nuclear Fusion 54 012002 (2014)

Austenitic stainless steels are widely used as structural materials for nuclear energy facilities. Investigations of microstructure and properties of reactor steels under irradiation is an important task for modern materials science. One of the examples of radiation-induced microstructural changes is related to the development of denuded zones along grain boundaries. This research aims to combine data from previous investigations on denuded zones and to present results obtained from studies of stainless steels (1^st and 2^nd generations) from the fuel assemblies of the decommissioned BN-350 nuclear reactor (Aktau, Kazakhstan). The studied materials of shrouds from the fuel assemblies were 12Cr18Ni10Ti (AISI321 analog) and 08Cr16Ni11Mo3 (AISI316 analog). The stainless steels were irradiated with neutrons to damage dozes of 0.25-59 dpa in the temperature range of 280-423°C and after the decommissioning they were kept in water environment for 4-8 years. The investigated samples were cut from shroud edges at a different distance from the core along the fuel assembly’s height. Transmission electron microscopy (JEOL JEM2100) was used to perform comparison studies of defect structure in the matrix, as well as that of grain boundaries and near-grain boundary regions. The effect of damage dose on the width of denuded zones and pores distribution was determined. EDS analysis gave information on elements distribution (in particular, Cr and Ni) at different distances from a selected grain boundary with a denuded zone. The work discusses factors that affect the development of denuded zones, it also provide quantitative results of defects sizes and redistribution of elements. Some obstacles of performing quantitative EDS analysis for the irradiated steel samples with swelling were discussed. As a result, the findings may help to illustrate microstructural changes taking place in the austenitic stainless steels after irradiation to further improvement of the existing structural materials for nuclear energy applications.

Fluoride salt-cooled High-temperature nuclear Reactors (FHRs) is emerging as a leading reactor concept among all Gen IV nuclear reactors because it offers, among other benefits, a high degree of passive safety, high thermal efficiency, and low spent fuel. One primary challenge in the development of the FHR is the selection of structural alloy that is required to be durable at 700°C, and to be compatible with molten Li₂BeF₄ (FLiBe) salt, as well as to be stable in strong neutron irradiation environment. Among many candidate alloys, nickel-based Hastelloy N^® and 316 stainless steel have been selected as the most promising structural alloys for FHR. In order to evaluate the performance of these two alloys in FLiBe salt in simulated environment of FHR, the static in-reactor corrosion tests in ⁷Li-enriched molten FLiBe salt at 700°C in both metal-lined graphite crucibles and bare graphite crucibles were successfully accomplished in MIT nuclear research reactor for 1000 hours. This study focuses on the post-irradiation examinations of these two in-reactor corrosion tested alloys using a suite of techniques, including scanning electron microscopy (SEM) in conjunction with energy dispersive spectroscopy (EDS), x-ray diffraction (XRD), transmission electron microscopy (TEM). From the microstructural results, the acceleration effects of neutron irradiation and graphite presence on the corrosion of these two alloys in molten FLiBe are understood which provides valuable reference to the materials selection for FHR.

Radiation-induced embrittlement of reactor pressure vessel is an important issue for the life extension of the nuclear power plant. Understanding the long-term evolution of its microstructure is a major challenge. In these steels, rich solute clusters formed under irradiation contribute to the mechanical properties evolution. According to tomographic atom probe analyses, these clusters are mainly enriched in Cu, Ni, Mn, Si and P. However, the elementary mechanisms leading to their formation are still an open question.

The formation mechanisms of these clusters are tackled thanks to multiscale modeling based on electronic structure calculations (DFT) and Atomic Kinetic Monte Carlo (AKMC). Defect and small cluster energetics are obtained by DFT allowing to develop a cohesive model used in AKMC to simulate diffusion and obtain the microstructure evolution. AKMC is very time consuming for a system in conditions close enough to those in reactor. Strategies used to speed up the simulations will be exposed. Furthermore, improvements of the AKMC parameterization based on recent ab initio calculations will be presented. High dose simulated microstructures will also be compared to experimental results.

Silicon carbide is a candidate material for the use in novel accident tolerant fuel cladding in nuclear technology due to its favorable properties, in particular reduced (compared to Zircaloy) oxidation under accident conditions, as well as good neutronic performance, high temperature strength and stability under irradiation. It is suggested to be used in the form of SiC-fiber reinforced SiC-matrix (SiC-SiC) composite.

Highly non-uniform and anisotropic nature of the composite materials means that in order to reliably model their behavior the knowledge of the individual properties of fiber and matrix, and, crucially, the fiber-matrix interfaces, is required. Micromechanical testing techniques, such as microcantilevers beam fracture, allow determination of such localized properties. This contribution for the first time reports the results of micromechanical measurements, coupled with microstructural characterization, on SiC-SiC composite material.

Material used in this study was provided by General Atomics. It consists of the commercially available Tyranno SiC fiber weaved reinforcement structure and matrix grown in-situ using the chemical vapour infiltration (CVI) technique. General structure of the composite, including fiber arrangement and porosity, was assessed using scanning electron microscopy (SEM) and X-ray tomography. Microstructure of fibers and matrix was characterized with electron backscatter diffraction (EBSD), transmission electron microscopy (TEM) and transmission Kikuchi diffraction (TKD) techniques. Fibres are found to be nanocrystalline with a grain size of ~50 – 300 nm. Microstructure of matrix is shown to be complex, with highly elongated grains (~200 – 500 nm wide, up to several microns long), arranged radially forming rings surrounding individual fibers.

Micromechanical studies included hardness measurements on fibers and matrix performed with nanoindentation, and interfacial fracture tests using focused ion beam (FIB) manufactured microcantilevers, both at room and elevated temperatures. The hardness of the fibres is shown to be significantly lower than that of the matrix (~25 and ~45 GPa respectively), this is thought to be due to porosity in the centre of the fibres. As a result of indentation in the fiber, extrusion of the carbon interlayer between fibre and matrix upwards form the surface was observed. Interrupted microscale interfacial fracture tests showed that failure in mode I fracture proceeds through the carbon interlayer, rather than down either SiC-C interface. Understanding how this interlayer fails is key to understanding the macroscopic properties of SiC-SiC composites, and presented results will be related to macro-scale behavior.

The influence of temperature and cooling rate on the embrittlement and the copper level in the matrix has been investigated on a weld fabricated from the same weld wire used for HSSI Weld 73W. This weld has a relatively high bulk copper content, 0.32% wt. The heat treatment consisted of heating the material to the desired temperature, holding at the post-weld heat treatment (PWHT) temperature, and then cooling down to room temperature. Except for special cases, all PWHTs were performed with a heating and cooling rate of 15^oF/h (8^oC/h) to simulate the heating/cooling rate of a real vessel. In two special cases, material was heated with 15^oF/h (8^oC/h) rate but water quenched after holding at the PWHT temperature. The highest PWHT temperature was 650^oC/24h, while the other PWHTs were 610^oC/24h, typical PWHT of reactor pressure vessels (RPV), and 580^oC/100h. Charpy impact properties were measured using sub-size 3x4 mm specimens and matrix Cu content was measured by atom probe tomography before and after irradiation. Small-angle X-ray scattering (SAXS) was used to measure size and volume fraction of copper-rich precipitates in the irradiated specimens. Charpy specimens were irradiated in the Ford Reactor at 288^oC to 0.8x10¹⁹ neutron/cm² (E>1MeV). It was found that the higher PWHT temperature resulted in higher Charpy upper-shelf energy (USE) with little effect on the ductile-to-brittle transition temperature (DBTT). The lower PWHT temperature and slower cooling rate were found to be beneficial in reducing the matrix Cu content. The matrix Cu content after irradiation was approximately the same for all three welds measured regardless of their different matrix Cu contents in the unirradiated condition. Consequently, the weld with the lowest PWHT temperature exhibited the lowest shift of DBTT and drop in USE. SAXS results reveal smallest volume fraction of copper-rich precipitates in this weld compared to others.

We have performed classical molecular dynamics simulations of swift heavy ion damage, typical of fission fragments, in nuclear fuel (UO₂) and surrogate material (CeO₂) with energy deposition per unit length in the range of 10-40 keV/nm. We did not observe amorphization. The damage consisted of isolated point defects at lower ernergies and defect clusters at higher energies. The simulations provide valuable insights on density changes in the ion track and the role of radiation damage in fission product diffusion. These results shed light on features observed by electron microscopy of swift heavy ion irradiated ceria.

The influence of microstructure on the mechanical properties of SiC layer coated onto spherical simulated nuclear fuel particles by fluidized-bed chemical vapor deposition (FBCVD) was investigated. While nano-indentation system was utilized to study mechanical properties of SiC layers, their microstructure were characterized by scanning electron microscopy (SEM) with EBSD, raman spectroscopy, and magic angle spinning nuclear magnetic resonance (MAS-NMR). The lower temperature deposition of SiC layer is advantageous factor preventing the property degradation of inner pyrolytic carbon layer which occurs at ~1500^oC. It was found that SiC layers deposited at lower temperature than 1400^oC obtained small grains sized between 0.2 and 1.0 µm and the non-stoichiometric ratio of Si and C. However, those phenomena were not observed in SiC layers deposited at higher temperature than 1400^oC. Young’s modulus and hardness measurements of SiC layers result in 302~358 and 32~40 GPa, respectively.

Liquid lithium (Li) and tin (Sn) are potential candidates for plasma facing materials in the divertor and first walls of fusion reactors due to attractive properties for self-recovery and heat-flux management. It is important to develop a strong understanding of how these metals are affected by the high-temperature environments that are typical in fusion devices. Surface science experiments under UHV conditions have proven to be a useful approach to study these materials and individual particle-surface interactions by enabling clean and controlled deposition of films and their subsequent behavior. We communicate recent investigations from UHV experiments on the thermal stability of ultrathin Li and Sn films on a polycrystalline W surface using temperature programmed desorption (TPD), Auger electron spectroscopy (AES) and low energy ion scattering (LEIS). This multitechnique approach provides a good picture of the various surface processes that occur such as diffusion, alloying, and desorption. Comparisons of the oxidation kinetics following exposure to residual gases on the solid and liquid films as obtained from oxygen uptake curves will be discussed. Additionally, the effect of both post-oxidation as well as preadsorbed surface oxygen and carbon on the thermal stability of these films will also be discussed. We also plan to report ongoing work to study deuterium uptake from incident beams of low energy D₂⁺ and atomic D on Sn and Li films as a function of temperature and film thickness using TPD.

The study presents results of observations and experiments on the effect of neutron irradiation on corrosion resistance of the structural materials from BN-350 fast nuclear reactor. In the process of decommissioning and utilization of the spent fuel assemblies (about 8 years of exploitation and storage in water environment) an observation of shroud’s surface was performed. Materials of the shrouds were 12Cr18Ni10Ti and 08Cr16Ni11Mo3 austenitic stainless steels. The calculated damage doses for the studied samples were in the range of 55-59 dpa, the irradiation temperature did not exceed 430°C. Visual and metallographic observations indicated the presence of corrosion-induced defects on shrouds surface. It was determined that the rate of corrosion was the most intensive at internal surface of the shrouds. There were also observed local defects, such as pitting, cavities, and inter-crystalline cracks which penetration into the material has reached about 45% of the shroud thickness. The performed analysis of the element composition let us to conclude that neutron irradiation resulted in concentration changes of Cr, Ni and Ti in certain areas of the sample, and it was assumed to be a cause for corrosion development. Laboratory examinations of the corrosion rate for the samples irradiated to different damage dozes showed that weight loss in the iron chloride increases in dozens of times with a damage doze increase. It was determined that irradiation with neutrons significantly changes steels microstructure, results in radiation-induced diffusion of alloying elements, development of alpha-martensite which decreases corrosion resistance of austenitic steels.

Martensitic α'-phase forming in austenitic stainless steels, the materials for fission and fusion reactors exposed to irradiation, stess and temperature gradients can sufficiently influence on both mechanical and corrosion properties that eventually reduces the operation period for reactor structural components. So it is actual to carry out the detailed investigation of peculiarities of α'-phase formation, growth and annealing in irradaited steels.
The present work reports the results of complex research of structure and properties changes for the reactor 18-10 chromium-nuckel steel caused by both the direct γ→α' and the reverse α'→γ martensite transformations after irradiation with high-energy charged particles (pulsed electron flux, ions of helium and krypton). Plain-shaped specimens for mechanical tests with gauge dimensions of 10×3.5×0.3 mm are comparatively investigated in both nonirradiated and irradiated states.
The “irradiation-induced martensite” was found in 12Cr18Ni10Ti steel directly after bombarding with a powerful pulse electron flux with energy of 150-300 keV and power density of 5×10⁸ – 4×10⁹ W/cm². . The martensitic nuclei forming in sites of a cellular dislocation network were observed in 15-20 mm depth from irradiation surface. On the opposite surface the lath–shaped martensite was observed forming in the result of shock compression in lattice during the puls electron flux action.
Experiments indicated that irradiation of 12Cr18Ni10Ti steel with high-energy krypton ions ⁸⁴Kr⁺¹⁴ with 1.56MeV (the DC-60 accelerator, Astana, RK) generated on the bombarded surface numerous carbides, nitrides, gas blisters as well as microinclusions of martensitic α (bcc) and ε (hcp)-phases. At fluences of (1 ÷ 4)×10¹⁵ ion/cm² the martensite can be revealed by means of EBSD-method (JSM-7500F microscope), at higher fluences of (6 ÷ 9)×10¹⁵ ion/cm² the standard methods of phase analysis as X-ray diffractometry and magnetic measurements can be used. It was found that an increase in fluence of ions resulted in an increase of α'-phase volume fraction forming in steel.
For studying the reverse martensitic transformations the specimens of 12Cr18Ni10Ti steel irradiated with helium ions (50 MeV) uniformly in volume (U-150M cyclotron, Almaty, RK) to 10^-3 at.% were deformed and subjected with a series of isochronous (30 min) annealings over the temperature range of 300-800^oC. As shown, the reverse a¢®g transformations did not change in monotonic way during the temperature increase, at temperatures of 400-500^oC there was an increase in a'-phase content caused by relaxation of compression strains in the steel lattice.

Zirconium based alloy is widely used as fuel cladding materials and some other structural materials in water-cooled reactors. However, degradation of zirconium alloy by waterside corrosion in harsh conditions severely limits the burnup and thus cycle life of nuclear fuels. Extensive researches have been done to understand this problem; however, several key aspects of knowledge related to the corrosion mechanisms of zirconium alloy still remain unclear. So far, most microstrutural analysis of corrosion behaviors of zirconium alloy is performed ex-situ. While few in-situ TEM studies of Zr-4 oxidation have been performed, none of them reaches atomic resolution, hindering the understanding of detailed corrosion process.

We performed in-situ environmental transmission electron microscope (E-TEM) study of zircaloy-4 initial oxidation in gas environments at atomic resolution for the first time. Experiments under water vapor and oxygen gas environments are done for comparison. Our preliminary analysis shown growth of nano-sized zirconium oxide crystals at the sample edge. This research will help discover new corrosion mechanisms of zirconium alloy, and thus provide insights on prediction and prevention of zirconium alloy corrosion in nuclear reactors.

"RBWR" （Resource- renewable BWR) is a new type of BWR which concentrate neutron flux to a specific part of the fuel and accelerate transmutation of TRU elements, in order to reduce the nuclear waste. The increased neutron flux may or may not accelerate the oxidization or the hydrogen pickup of the Zircalloy fuel claddings. The precise effect of the irradiation on these degradation phenomena is still unclear, and we must clarify whether or not the increased flux shortens the service life of the claddings. Based on the fact that irradiation induced strain breakaway and hydrogen pickup start at similar dose, we speculate that c-type vacancy loops that induce the strain breakaway are responsible for the nucleation of hydride. Using first-principles calculations, we show that basal surface can be saturated with hydrogen with a binding energy of 0.5eV, and two hydrogen-filled surfaces can close in to a distance only 15% greater than the distance between two successive basal layers in bulk Zr. Closed-in surfaces have a similar atomic structure as ZrH2 and promote further hydrogen absorption to the adjacent layers, acting as a nucleation site for the hydride.

Ferritic steels based on Fe-Cr system are perspective structural materials for fusion applications. Helium is produced in the structural materials in nuclear power plants by the nuclear transmutation following the neutron irradiation. Since the solubility of helium in all the metals is extremely low helium tends to be trapped at defects such as vacancies, dislocations and grain boundaries, which causes materials embrittlement. Helium atoms influence both a cohesive energy and magnetic properties of grain boundaries in Fe-based alloys. In these studies, the segregation energy of helium at symmetric Σ3 (111) and Σ5 (210) tilt grain boundaries (GBs) is investigated both in Fe dilute alloys and Fe-Cr alloys as well as in the Fe-Cr structures generated using DFT-based Monte Carlo simulations [1]. In particular, the strengthening / embrittlement energies caused by He and the segregation energies of He are calculated for different concentrations of Cr and He atoms. The results show that the presence of helium atoms strongly influences the magnetic properties of system in the relatively distant neighbourhood. It is observed that the segregation energy of helium atom is influenced by Cr location in GBs. The migration paths and the migration energy barriers of He atoms in GB structures of Fe-Cr alloys are calculated using the Nudged Elastic Band method. Density functional theory calculations are performed by using VASP code, with generalized gradient approximation (GGA) of Perdew-Burke-Ernzerhof (PBE) for exchange-correlation.
The modelling on the atomistic level presented here will enable to improve the knowledge on the phenomena of helium embrittlement and formation of helium bubbles in Fe-based alloys.
[1] J. S. Wróbel, D. Nguyen-Manh, M. Yu. Lavrentiev, M. Muzyk, S. L. Dudarev, Phys. Rev. B 91, 024108 (2015).

As an advanced oxide dispersion strengthened iron alloy developed through mechanical alloying, 14YWT has exhibited excellent mechanical properties under high temperature, high pressure and high irradiation condition. Other than nanometer size precipitants such as TiN, TiC, and Y₂Ti₂O₇, there exists a high density (>10²³) of ultra-fine Y-Ti-O enriched nanoclusters (2~4nm in diameters) within the iron matrix. These oxygen enriched nanoclusters are ultra-stable without coarsening at high temperature, high pressure and high irradiation conditions. After ion/neutron irradiation, the helium bubbles are observed extremely uniform in size (1nm) and quite homogeneously distributed within the matrix and next to these nanoparticles indicating that the microstructure of 14YWT remains remarkably tolerance to radiation damage. With first principles theory calculation, we will perform the energetic study about the helium interaction with different local environments in 14YWT, such as (1) helium atom in iron matrix with/without vacancy; (2) helium atom within a BCC iron unit cell enriched with Y, Ti and O:vacancy pair; (3) helium atom within TiN/TiC matrix with/without vacancy; and (4) helium atom at the coherent/semi-coherent Fe-TiN/TiC interface enriched with vacancy and O:vacancy pair. The objective is to define the preference locations where helium bubbles would like to nucleate, therefore to further understand the formation and growth criteria of helium bubbles within 14YWT.

In recent years, the phase field approach has become a popular and powerful tool for modeling nuclear materials, including UO₂ fuel. Such models are able to efficiently predict fuel properties such as microstructure, heat conductivity, fission-gas release rate, and other properties. However, current fuel-performance phase field codes are often limited by assumptions of the fuel’s initial microstructure. One way to improve the initial microstructures of these models is to develop a phase field model of the fuel manufacturing process. The result of the manufacturing model could then be used as the initial condition of the fuel-performance model.

UO₂ fuel pellets are manufactured by sintering to densities of 95% or greater. The remaining porosity has a large impact on fuel performance. Once the fuel is placed in a reactor, it continues to condense due to heat, pressure, and irradiation effects. The current work seeks to build on existing work to develop a mechanistic phase field model that describes densification of UO₂ both during sintering and during reactor operation in the MARMOT mesoscale fuel performance code. In this presentation we will illustrate how the model has been implemented in MARMOT and demonstrate how it extends the capabilities of MARMOT to model fuel fabrication.

A quantum mechanical/molecular mechanical method was employed in the investigation of interactions between impurity atoms and dislocation cores (both screw and edge) in bcc materials, especially iron and tungsten. Hydrogen atoms are particularly considered for tungsten, while Cr, Ni and Cu atoms are considered for iron. The configurations of the dislocation core structures are the results of two competing energies: the interaction between the fractional dislocations and the corresponding generalized stacking fault energy in between the two partial dislocations, which is presented in this work. With this, we can precisely predict the configurations of the interstitial/substitute doped dislocation core structures, which are of significant importance for the future understanding of the dislocation mobile properties.

Grain refinement has been proposed as a promising way to improve the radiation resistance of silicon carbide (SiC), as grain boundaries are known to act as defect sinks. However, the detailed mechanisms for defect annealing at grain boundaries still remain unclear. Characterizing the radiation-induced composition change of grain boundaries can provide important clues for resolving the complicated process of defect-grain boundary interactions. We applied electron energy loss spectroscopy (EELS) to characterize the grain boundary chemistry in CVD (chemical vapor deposition) 3C-SiC bulk samples at different irradiation conditions. The experiments found grain boundaries in CVD 3C-SiC are intrinsically carbon-poor within 2 nm nearby grain boundaries with a relative C composition as low as 45%. This carbon depletion disappeared in samples irradiated to 1 dpa (displacement per atom) at 300 C, which is probably due to the unbalanced flux of C interstitials to defect sinks and low diffusivity of defects at grain boundaries. The C depletion appeared again in samples irradiated at 600 C and the depletion is obviously larger (in about 4nm-wide region near grain boudaries and the relative C composition can be as low as 35%) than that in the non-irradiated sample. This indicates that the role of grain boundaries may change from defect clustering reservoirs at low irradiation temperature to defect diffusion channels at high irradiation temperature. This transition has also been predicted by our ab initio informed rate theory model

The low thermal conductivity of UO₂ ceramics causes large thermal gradients in nuclear fuel pellets, which in turn results in degradation of the fuel from thermal stress induced fracturing and other defects. With this in mind, more accident tolerant reactor fuel is being developed that employs additives to raise the thermal conductivity of UO₂ fuel. In this work, we employ the MARMOT mesoscale fuel performance code to evaluate the impact of BeO and SiC additives on the effective thermal conductivity of the fuel. These mesoscale simulation results are then used to inform the development of a thermal resistance model that predicts the effective thermal conductivity as a function of the additive volume fraction and average size. This model can be used in fuel performance codes such as BISON to evaluate the fuel performance of these accident tolerant fuel concepts.

We present phase-field simulations of microstructure evolution of metallic Uranium compacts exposed to sintering temperatures. The open source (M)ultiphysics (O)bject-(O)riented (S)imulation (E)nvironment software package is used to perform phase-field simulations of a green compact with varying initital distributions of particle size. The effects of the initial morphology is investigated for the purpose of determining configurations that allow highly stable pore distributions that resist densification and act as capture sites for fission gases. In particular, the sintering behavior of microstructures containing bimodal pore size distributions are studied and found to possess greater resistance to densification relative to monomodal pore sizes. Applications to metallic reactor fuels, prepared by powder metallurgy, that capture fission gases
and improve in-pile swelling are targeted.

Near-surface region of plasma facing material (PFM) plays an important role in thermal management of fusion reactors. In this work, we measured thermal conductivity of tungsten (W) surface layer damaged by He plasma in PISCES at UCSD and Cu²⁺ ion beam at LANL.

For He plasma irradiated W, we studied the damage effect on both bulk and thin film of W. The surface morphology of both bulk and thin film was altered after exposure to He plasma with the fluence of 1×10²⁶ m^-2 (bulk) and 2×10²⁴ m^-2 (thin film). Transmission electron microscopy (TEM) analysis reveals that the depth of the irradiation damaged layer was approximately 20 nm on a bulk W exposed to He plasma at 500 ^oC for 2000 s (with the fluence of 1×10²⁶ m^-2). In order to measure the thermal conductivity of this exceedingly thin damaged layer in the bulk W, we adopted the well-established ‘3-omega’ method and employed novel nanofabrication techniques to improve the measurement sensitivity. For damaged W thin film sample, we measured the reduction in electrical conductivity and used the Wiedemann-Franz (W-F) to extract the thermal conductivity. Results from both measurements showed that thermal conductivity in the damaged layers was reduced by at least ~80% compared to that of undamaged W.

Moreover, Cu²⁺ ion irradiation with low power density is performed on W to mimic neutron displacement damage. The thermal properties of the damaged layer (~1mm) were measured by the aforementioned adopted 3-omega method. Results show that the thermal conductivity of irradiated surfaces drops from the un-irradiated value of 182±3.3 Wm^-1K^-1 to 37 ±2.8 Wm^-1K^-1 for 2dpa, 50±7.8 Wm^-1K^-1 for 0.6 dpa, 52.5±8 Wm^-1K^-1 for 0.2 dpa irradiated at 500 ^oC respectively. For samples damaged at room temperature, decrease in thermal conductivity was observed even at 10^-3 dpa. The decrease appears to begin to saturate at a value of roughly 55 W/m-K as the damage level exceeds 10^-1 dpa, a factor of ~3x lower than the undamaged value.

The large decrease in thermal conductivity can be attributed to the scattering of electrons, the dominant heat carriers in W, caused by defects created during the He and Cu²⁺ ion bombardment process. Our result suggests that suppressed thermal conductivity in PFM needs to be taken into account in the thermal design of future plasma-facing components.

In order to prevent the hydrogen explosion induced radioactive isotopes contamination happened in the Fukushima Daiichi Nuclear Power Plant, we need to design new cladding materials having the following key properties: a reasonably low thermal neutron absorption cross section; a high resistance to radiation damage up to 10 dpa without degradation in property during normal operation; and an excellent high temperature oxidation resistance up to 1000 celsius with a reasonable mechanical strength during the accident conditions. Recent research work has been toward the design of high entropy alloys [1,2] by mixing together the various elements at or near equal molar proportions. Because of this "high-entropy effect", a multi-component alloy is a very stable composition at high temperature. High entropy alloy as a new type of multi-component alloy, not only has a very high intensity (> 1 GPa), but also has excellent toughness (fracture strain of 70%), that is, ultra-high resistance to injury (fracture toughness K_IIC＞ 200 MPam^1/2). As a result, these multicomponent alloys which have a high strength [1,3,4], sluggish diffusion [5,6], and excellent creep resistance [7], so that they become promising new materials [8,9] at high temperature.
HEAs’ radiation behavior, however, is scarcely studied although they are potential radiation resistant materials indicated by two intrinsic properties: (a) severe distorted lattice sites in HEAs can trap atoms which may lead to diffusion-limited interstitial-vacancy recombination at high irradiation temperature; (b) low stacking fault energy in some HEAs favors the formation of nanotwins which act as sinks inside each grain to assist interstitial-vacancy recombination too. Thus, swelling in HEAs may be controlled by tuning these two properties to reach an ideal balance of production and annihilation of interstitial and vacancy defects under irradiation, resulting in “self-healing” effect. Atomistic simulation is performed to study radiation defect clusters in HEAs, resulting crystal structure would be detailed analyzed. This investigation will lead to more understandings about the possibility of utilizing HEAs as next generation accident tolerance cladding material.

References
[1] Zhang Y, Zuo TT, Tang Z, Gao MC, Dahmen KA, Liaw PK, et al. Prog Mater Sci 2014;61:1e93.
[2] Yeh JW, Chen SK, Lin SJ, Gan JY, Chin TS, Shun TT, et al. Adv Eng Mater 2004;6:299e303.
[3] He JY, Liu WH, Wang H, Wu Y, Liu XJ, Nieh TG, et al. Acta Mater 2014;62:105e13.
[4] Zou Y, Maiti S, Steurer W, Spolenak R. Acta Mater 2014;65:85e97.
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Irradiation of crystalline structure in conjunction with stress causes the materials to slowly deform (“creep”), resulting in changes to the material’s dimensions and strength. One of the options to limit creep in coarse grained materials is to decrease grain size down to sizes where dislocation mobility is prevented and point defect sinks are abundant. It was previously demonstrated that in such systems creep is controlled by local “string” like atomic relaxations within the grain boundaries as they absorb point defects, e.g., interstitials and vacancies, produced in the grain interior during irradiation. We demonstrate that in such systems grain boundary impurities have strong effect on observed radiation induced creep rates while having insignificant effect on thermal relaxations. Characterization of creep related relaxations within the grain boundary and correlation between these relaxations for various immiscible impurities serve as a basis for creating a mean field model for irradiation induced creep. The model predictions are used to describe various segregation scenarios and sample structures.

[1] Ashkenazy, Yinon, and Robert S. Averback, "Irradiation Induced Grain Boundary Flow—A New Creep Mechanism at the Nanoscale." Nano Letters 12(8): 4084–4089

Equations describing the elastic interaction between dislocation segments are fundamental to discrete dislocation dynamics simulations. However, the typical level of spatial resolution of such simulations is of the order of a hundred Burgers vectors, and this impedes application of discrete dislocation dynamics to modelling interactions between radiation defects that have much smaller size. We derive expressions for the energy of elastic interaction between nano-scale defects, such as dislocation loops and vacancy clusters, produced in materials by irradiation. These expressions are well suited to real-space simulations of the dynamical evolution of many such interacting defects in an elastic continuum. In these simulations the elastic interaction energy between defects is analogous to the atomic interaction energy in a molecular dynamics simulation. This work was part-funded by the RCUK Energy Programme [grant number EP/I501045]. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

Irradiation embrittlement of reactor pressure vessel (RPV) steels could limit the lifetime extension of light water reactors (LWRs). Significant embrittlement can be caused by the formation of nanometer-scale Mn-Ni-Si precipitates (MNSPs). Recently, atom probe tomography (APT) experiments were used to investigate the thermodynamic stability of MNSPs under post-irradiation annealing (PIA). Here a kinetic Monte Carlo (KMC) model parameterized using CALPHAD and recent APT data has been used to simulate PIA of low-Cu RPV steels. The model predicts an evolution of the average size, number density, and volume fraction of MNSPs that agrees well with the APT data. The model also shows that during both the growth and dissolution of the MNSPs they deplete Si from their outer shell compared to the composition seen at the start of the experiment, a result consistent with the APT observations. It is proposed that due to the lack of mobility inside the MNSPs, diffusion limited-growth of MNSPs can lead to temperature dependent precipitate composition. The KMC model has also been used to investigate the morphology of the Cu-MNSP precipitates formed during irradiation and annealing. A kinetically limited growth mechanism is proposed to explain the fact that MNSP forms as an appendage on Cu clusters, rather than a shell coating the Cu core.

Recent experimental observations show the formation of chromium clusters in ion-irradiated dilute Fe-Cr alloys in the Cr concentration range close to 5 at. %. This is well below the 10 at. % Cr solubility threshold above which the chromium-rich phase is thermodynamically stable. The origin of this anomaly is unclear, however the fact that it occurs under irradiation suggests that vacancies and small vacancy clusters might facilitate Cr segregation, in this way changing the alloy performance under high temperature and high irradiation dose conditions. We have investigated the possible occurrence of chromium segregation near vacancies in dilute Fe-Cr alloys using ab initio calculations. We find that both single vacancies and vacancy clusters can serve as nucleation centres for chromium atoms. A single vacancy can attract up to five Cr atoms, with the most stable configuration corresponding to Cr atoms arranged as third nearest neighbours. Configurations involving second nearest neighbour Cr atoms are energetically unfavourable and have negative binding energies. A linear fit to the numbers of second, third and fifth nearest neighbours describes the binding energy of chromium-vacancy clusters very well. Clusters containing up to four vacancies and up to four Cr atoms were investigated and positive binding energies were also found. Substantial deviations of atoms from ideal crystallographic positions were found in larger clusters, suggesting that bcc structure may not be retained in the small cluster size limit. It was found that the absolute values of magnetic moments of Cr atoms in the energetically favourable configurations are substantially larger than in the unfavourable configurations, highlighting the magnetic nature of the anomalous Cr-vacancy segregation effect. In summary, we predict agglomeration of Cr atoms on vacancies and vacancy clusters in technologically important Fe-Cr alloys with small chromium content.

The objective of the Department of Energy, Office of Nuclear Energy Light Water Reactor Sustainability Program (LWRSP) is to support the safe operation of nuclear reactors to extended lifetimes of up to 80 years. One of the primary reactor components of concern is the reactor pressure vessel (RPV), which becomes increasingly embrittled over time due to damage from high energy neutrons leaking from the reactor core. Radiation damage leads to the formation of nm-scale precipitates and solute-defect clusters, which result in hardening and a corresponding increase in the steel’s ductile to brittle transition temperature.

Extended life operation will subject some RPVs to higher neutron fluences than previously experienced by in-service vessels. Thus, robust prediction methods must be developed that ensure the RPV safety for our nation’s fleet of reactors. Unfortunately, current regulatory models under predict embrittlement data from high flux test reactor irradiations at extended life fluences. It has been hypothesized that a major reason for this under prediction is the formation of so-called “Late-Blooming Phase” Mn-Ni-Si precipitates, which grow only at high fluence, and that are currently not included in regulatory models. Thus, a large-scale irradiation experiment was designed to provide a database needed to develop and validate extended life embrittlement prediction models. The irradiation was recently completed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory as part of the DOE sponsored Nuclear Scientific User Facilities program. The UCSB ATR-2 irradiation contained ≈ 1600 specimens composed of 172 alloys (including surveillance steels) in different specimen configurations. The fluence extended up to 1.4x10²⁰ n/cm² over range of temperatures from ≈ 250 to 310°C. A very large body of post irradiation examination results will be presented detailing the individual and combined effects of neutron flux, fluence, irradiation temperature and alloy chemistry on precipitate formation, hardening and embrittlement. In addition, a new high fluence embrittlement prediction model will be proposed.

Steels are among the most important structural materials in nuclear energy engineering. The only reactor component that is irreplaceable is the reactor tank, normally made out of bainitic steel. Here, we will present and review multi-scale modelling approaches for studying radiation damage effects: from the primary damage, basic interactions and transport properties based on electronic structure theory, to dynamic studies of displacement cascades and dislocation-defect interactions, to stochastic- and mean-field techniques that drive the
modelling into higher scales in space and time. We will discuss several effects of the radiation-driven microstructure evolution, including embrittlement, swelling and
segregation; as well as recent model development. We will present results including the effects of many of the major solute and impurity atoms in the steels and discuss how the modeling compares to and can aid in analysing experimental results.

Vacancy-mediated climb models cannot account for the fast, direct coalescence of dislocation loops seen experimentally. An alternative mechanism, self climb, allows prismatic dislocation loops to move away from their glide surface via pipe diffusion around the loop perimeter, independent of any vacancy atmosphere. Despite the known importance of self climb, theoretical models require a typically unknown activation energy, hindering implementation in materials modeling. Here, exten- sive molecular statics calculations of pipe diffusion processes around irregular prismatic loops are used to map the energy landscape for self climb in iron and tungsten, finding a simple, material independent energy model after normalizing by the vacancy migration barrier. Kinetic Monte Carlo simulations yield a self climb activation energy of 2 (2.5) times the vacancy migration barrier for 1/2<111> (<100>) dislocation loops. Dislocation dynamics allowing self climb and glide show quanti- tative agreement with transmission electron microscopy observations of climbing prismatic loops in iron and tungsten, confirming that this novel form of vacancy-free climb is many orders of magnitude faster than traditional climb models. We show that self climb significantly influences the coarsening rate of defect networks, with important implications for post-irradiation annealing.

The operation of existing fission reactors provides more than 10% electricity worldwide, nearly 20% in the US. The advance to Generation IV nuclear reactor brings more challenges in structural materials. To develop new materials that maintain their desired properties while being driven very far from equilibrium, we examined the radiation stability of amorphous silicon oxycarbide (SiOC) and crystalline Fe nanocomposites. The highlight of our previous work shows that amorphous SiOC alloys are thermal and irradiation stable over a wide range of composition, irradiation dose and irradiation temperature. Nanocomposites of these materials with crystalline Fe also exhibited good irradiation stability at elevated temperatures. More specifically, the swelling resistance in the Fe layers of SiOC/Fe (80/60 nm) multilayer film is improved by 2.2 times and the averaged void size is reduced to half compared to that of pure Fe films. In some instances a crystalline Fe_xSi_yO_z reaction layer formed between the Fe and SiOC components of the composite, and the interface between Fe and Fe_xSi_yO_z was observed to be incoherent. Void denuded zones were observed in the Fe layer close to the SiOC/Fe and Fe/Fe_xSi_yO_z interfaces. All the findings strongly support the idea that crystalline/amorphous SiOC/Fe interfaces and Fe/Fe_xSi_yO_z incoherent interfaces act as efficient defects sinks, extending the low-swelling transient regime and suppressing void swelling during irradiation at 600 °C.

This work presents experimental and modelling investigations on oxidation and thermal shock behavior of Silicon carbide (SiC) cladding for light water reactors (LWRs). Experimental investigations are conducted on sintered tubular SiC in oxidizing environments containing pure steam at 1 atm with temperature range of 1140–1500 °C and velocity between 0.8 and 10 m/s. Linear weight loss was observed with time. The linear weight loss rates exhibit sensitive dependence on flow rate at a given temperature, demonstrating effects of flow boundary-layer diffusion rate on silica volatilization kinetics. Silica scale exhibits morphology change with respect to exposure time in an oxidizing environment, progressively demonstrating bubble formation and surface smoothing, extensive formation of cracks and pores, and crack reduction. Yet, strength measurements of pressure-less sintered SiC show no significant change after oxidation in the tested conditions. Correlations for weight loss rate of pressureless sintered SiC with steam flow velocity, and temperature in the tested conditions were developed. While preserving physical significance of non-dimensional numbers dictating boundary layer mass transfer phenomena, the correlations capture boundary layer diffusion effects on the silica volatilization rate, which is affected by flow characteristics at the oxidizing surface. SiC oxidation rates were shown to be about three orders of magnitude lower than the oxidation rates of Zircaloy-4. High purity CVD-SiC (1.5 mm x 2.0 mm in thickness) exhibits thermal shock tolerance~1260 ^oC in room temperature water and beyond it (>1260^oC) in saturated water. Decreasing water subcooling significantly reduces thermal shock fracture danger of ceramic materials. Solid-liquid contacts during nucleate and transition boiling, and boiling incipience upon water bath entering are a highly probable origin of thermal shock fracture in water quenching. No single constant heat transfer could suffice to depict the actual stress evolution subject to dynamic fluid phase changes. Use of the surface temperature dependent heat transfer coefficient will remarkably increase predictability of thermal shock fracture of brittle materials.

Silicon Carbide and Silicon Carbide composites are being studied as potential materials in multiple nuclear applications, from TRISO-coated gas-reactor fuel particles to flow channel inserts in the Pb/Li breeder blanket in fusion rectors or other first wall and breeder blanket applications in other fusion devices. Data on the effects of irradiation on the mechanical behavior of SiC remains sparse, and is needed for development of these applications.
In this work ion implantation is used as an analogue for neutron irradiation due to the reduced costs and campaign times required for ion implantations. A sequential silicon ion implantation was performed at the Surrey Ion Beam facility on commercially available SiC (Direct Sintered SiC). Samples were irradiated with 6x10¹⁵ ions/cm² at 2.0 MeV, 4x10¹⁵ ions/cm² at 1.0 MeV and 2x10¹⁵ ions/cm² at 0.5 MeV at 300^°C. This gave a peak dose of ~3.8 dpa with damage to a depth of ~1.4μm. Due to the small damage layer micro-mechanical testing techniques were used to look at the post irradiation mechanical properties and ramen spectroscopy and transmission electron microscopy to study the effects of irradiation on the microstructure.
Nanoindentations were performed to a depth of 1 micron in both implanted and unimplanted material. The damage layer was seen to harden by approx. 4 GPa, with an increase from 38.9±2.5 GPa to 43.3±1.6 GPa. This effect is consistent with radiation damage in metals. The elastic modulus however, was seen to decrease before tending back to the bulk value, reducing from 471±24 GPa in the unirradiated state to 431±16 GPa in the irradiated state. This effect is consistent with the literature and thought to be due to lattice swelling, and is a clear example of the differences between effects of irradiation on metals and ceramics. The Ramen spectroscopy and TEM showed that the damaged layer had become amorphous during irradiation. The implications for this on the use silicon carbide for nuclear applications in both fusion and fission reactors will be discussed.

Nanostructured ferritic alloy (NFA) and silicon carbide (SiC) are two promising cladding materials which have received extensive attention due to their mechanical stability and irradiation resistance. NFA-SiC composites are expected to have enhanced performance by combining the merits of each composition for nuclear applications. Spark plasma sintered pure NFA (100 vol% NFA-0 vol% SiC) and NFA-SiC composites (97.5 vol% NFA-2.5 vol% SiC, 95 vol% NFA-5 vol% SiC) with comparable mechanical strength have been achieved in our previous works. Yet, the harsh environment of high temperature moist condition is a critical factor that affects the long term operation of nuclear cladding materials. In this report, we have systematically studied the effect of water vapor thermal treatment on the microstructure evolution of the pure NFA and NFA-SiC composites from 500-1000^oC. The phase compositions of surface oxidation layers are investigated to be comprised of Fe₂O₃ and Fe₃O₄, while the grain growth of Fe₂O₃ along (300) and (006) planes becomes predominant with the increase of the thermal treatment temperature. The oxidation layer evolves from flake structures at 500^oC, to deformed roundish structures at 750^oC, and to stub structures at 1000^oC. The thickness of the oxidation layer is determined by the composite composition, the composite density, and the thermal treatment temperature. The much improved oxidation resistance is further confirmed in the NFA-SiC composites by forming a dense SiO₂ layer which acts as oxygen diffusion barrier on the surface. The NFA-SiC composites are promising corrosion resistant materials for nuclear applications under harsh high temperature moist environment.

This paper introduces on-going programs on “Increased Accident Tolerance of Fuel for LWRs”, where SiC/SiC R & D is the key activity including reactor irradiation tasks. MEXT fund based project, named as SCARLET and METI fund based projects, named as INSPIRE have been launched as 5 year termed projects at OASIS in Muroran Institute of Technology since 2012. These projects aim to replace Zircaloy claddings to SiC/SiC based materials.
Those programs are developing large scale production of all SiC/SiC cladding with sufficient gas leak tightness and also developing technology integration to make fuel pin model segments.
In-reactor irradiations of SiC/SiC with SIC/SiC or Zircaloy end-caps have been completed under dynamic reactor water environment with water chemistry control in Halden, Norway.
The results of Halden reactor irradiation, out of reactor core test results and the current status of large scale SiC/SiC cladding production status will be presented.

Gilsocarbon graphite is a structural core component in the advanced gas cooled reactors (AGRs) that supply approximately 15% of UK’s electricity. The polygranular graphite components moderate the fast neutrons for efficient fission, guide the flow of coolant gases and provide channels for the movement of fuel and control rods. Graphite is also a critical material for several designs of future Gen IV reactors that will operate at much higher temperatures; these include the U-battery gas cooled fast reactor, which is a UK designed micro-reactor (<10 MWe) for distributed power generation. In these applications, for safety reasons, it is essential to assure the structural integrity of the graphite components. This requires a sound understanding of the structure-property relations, since fast neutron irradiation (and radiolytic oxidation in the case of AGRs) can have significant effects on graphite properties.
Polygranular graphites have complex multi-scale microstructures. The Gilsocarbon graphite comprises a matrix with coarse filler particles and has porosity ranging in size from nano-scale to several hundred micrometres. In situ tests using diffraction are needed to provide a “bridge” between the nano-scale characteristics of graphite, as exemplified by the lattice strains in its network structure of fine crystals, and the macro-scale properties of the heterogeneous microstructure. This allows quantification, under applied load, of the relationships between the elastic strains (i.e. stress) and inelastic deformation (i.e. total strains).
Our previous room temperature study used neutron diffraction to explore the deformation of Gilsocarbon nuclear graphite under compression and tension via four-point bending at ENGIN-X, ISIS (Rutherford Appleton Laboratory, UK). This work is the first such investigation at elevated temperatures, typical of reactor operation (650C and above). ‘Brazilian’ disc Gilsocarbon graphite specimens were loaded under diametric compression, to generate a biaxial tension/compression stress state. A halogen lamp-heated furnace with an Argon gas environment to prevent oxidation of graphite was combined with an in situ loading cell (both designed at ENGIN-X) and high temperature optical digital image correlation of the specimen under load. This allowed simultaneous measurement of diffraction spectra to examine the effects of the orthogonal tensile and compressive stresses, and also the bulk compressive deformation of the specimen.
Results are reported for tests of specimens loaded to failure at room temperature and 650C; these will be discussed in terms of the lattice thermal expansion with temperature in the stress-free condition, and the correlation between lattice strain and applied stresses at temperature, and the effects of temperature on the stress/strain behaviour of virgin Gilsocarbon graphite at elevated temperatures. The potential for testing irradiated graphites at service temperature is discussed.

Nuclear-fission and fusion materials are degraded primarily due to the accumulation of lattice defects in response to energetic neutron irradiation, where point-defect clusters are directly produced in addition to point defects (self-interstitial atoms (SIAs) and vacancies) in the primary damage process, which is called “collision cascade.” To precisely predict the life time of these materials, accurate understanding of the structures and dynamic properties of radiation-produced defects is required.

We have directly revealed dynamic properties of individual nanoscale SIA clusters in metals, using in-situ TEM ([1-3]). In addition, we have extracted the dynamic properties of TEM-invisible single SIAs, from the formation process of TEM-visible SIA clusters under high-energy electron irradiation, where primary defects generated via knock-on displacement are only point defects unlike neutron and ion irradiation, with a high-voltage electron microscope (HVEM) ([4, 5]).

In this presentation, we will primarily focus on the structures and dynamics of radiation-produced point defects and point-defect clusters in high-purity tungsten. Firstly we will show some of our results on dynamic properties of SIAs and SIA clusters, which have been mainly obtained with HVEMs in Osaka University and Nagoya University in Japan. And, we will provide our results on the formation processes of point-defect clusters under self-ion irradiation, which have been obtained with an ion-accelerators combined TEM in the JANNuS-Orsay facility in France. Through the comparison between these results, we will try to extract the effects of collision cascades in the formation processes of point-defect clusters.

References
[1] K. Arakawa et al. Phys. Rev. Lett., 96 (2006) 125506.
[2] K. Arakawa et al. Science, 318 (2007) 956.
[3] K. Arakawa et al. Acta Materialia, 59 (2011) 141.
[4] T. Amino, K. Arakawa, and H. Mori: Philos. Mag. Lett., 91 (2011) 86.
[5] T. Amino, K. Arakawa, and H. Mori, Scientific Reports, 6 (2016) 26099 .

Recent developments of advanced transmission electron microscopes allow us to observe the detailed structures of individual atomic defects during formation and evolution. We show here a series of structural evolutions of atomic defects triggered by high-energy electron beam. First example is graphene with various types of defects. Especially formation and thermal relaxation of interstitial-vacancy pair defects between two graphene layers was successfully observed and tried to correlate to the Wigner energy release [1]. The other examples for defect evolutions of doped graphene or transition metal dichalcogenides will be also shown [2, 3]. [1] K. Urita et al. Phys Rev. Lett. 94, 155502 (2005) [2] YC. Lin et al., NanoLett. 15 (2015) pp.7408-7413 [3] YC. Lin et al., Nature Comm. 6 7736 (2015) [3] This work is supported by JST-CREST and JST-ACCEL research programs.

Irradiation can lead to a significant change in deformation mode in common structural materials used in nuclear reactors. In this study, the deformation behavior of unirradiated and neutron-irradiated 316 stainless steel samples were characterized in situ with wide-angle X-ray scattering and small-angle X-ray scattering at the Advanced Photon Source. The samples were solution-annealed 316 SS irradiated to 10 dpa at 320°C. While the unirradiated specimen showed uniform macroscopic deformation, the irradiated sample exhibited a localized plastic flow, initiating from one end of the tensile gauge and propagating towards the other end. In addition, martensitic phase transformation was observed in both unirradiated and irradiated samples during deformation, and the phase transformation was directly linked with plastic deformation. Full-gauge X-ray diffraction scans at several intermittent deformation levels showed microstrain partitioning on different lattice planes in different phases, revealing the operating deformation mechanisms in the irradiated material. The deformed sample was further investigated using the high-energy X-ray diffraction microscopy technique (also known as 3D-XRD), which revealed the refinement of orientation domains and the distribution of residual strain/stress. The combination of the in-situ and the 3D X-ray studies sheds new light on the irradiation-induced deformation mode change in structural materials.

Metallic alloys used as structural materials in the nuclear core of pressurized water reactors undergo irradiation creep. A proper understanding of the mechanisms which control the deformation process is essential in order to predict the dimensional changes under irradiation.
A novel approach based on the testing of on-chip thin freestanding structures [1-3] has been for the first time extended to be applied in the context of irradiation studies [4]. An elementary test structure is composed of three main elements: (i) a thin specimen of the material of interest, (ii) an actuator layer made of silicon nitride with strong internal tensile stresses to deform the attached specimen layer and (iii) a sacrificial layer to release the test structures from the underlying substrate. The small thickness of the material, below a few hundred nanometers, allows full and almost homogenous irradiation by heavy ions with a kinetic energy of a few hundred keV.
After adapting the method to in situ irradiation conditions, several sets of structures were successfully tested to assess the room temperature creep behaviour of copper films produced by physical vapour deposition. The contribution of the irradiation creep to the deformation has been quantified by performing several irradiation steps up to a dose of 8 dpa and measuring after each step the deformation of the specimens.
The results show that the creep under irradiation at room temperature is significant. The creep rate follows a power law dependence with the stress.
In parallel to these nanomechanical tests, the microstructure of the copper film before and after irradiation has been characterized by transmission electron microscopy. In particular, the influence of the grain size on the irradiation defects has been investigated. No significant effect on the irradiation defects is found for grain size between 100 nm and 10 µm.
References
Gravier, S., Coulombier, M., Safi, A., Andre, N., Boe, A., Raskin, J.P., Pardoen, T., 2009. New MEMS – Based micromechanical testing laboratory – Application to aluminium, polysilicon and silicon nitride. Journal of Microelectromechanical Systems 18, 555-565
M. Coulombier, G. Guisbiers, M.-S. Colla, J.-P. Raskin, T. Pardoen, On-chip stress relaxation testing method for freestanding thin film materials, Review of Scientific Instruments 83 (2012) 105004-1 to 9
M.S. Colla, B. Amin-Ahmadi, H. Idrissi, L. Malet, S. Godet, J.P. Raskin, D. Schryvers, T. Pardoen, Dislocation mediated relaxation in nanograined columnar palladium films revealed by on-chip time-resolved HRTEM testing, Nat. Commun. 6:5922 6922 (2015)
P. Lapouge, F. Onimus, R. Vayrette, J.-P. Raskin, T. Pardoen, Y. Bréchet, A novel on chip test method to characterize the creep behaviour of metallic layers under heavy ion irradiation, Journal of Nuclear Materials 476 (2016) 20-29

Post-irradiation annealing (PIA) is an useful tool to deconvolute the microstructural contributions to the performance degradation of irradiated stainless steels. It is also potentially an effective mitigation strategy. Literature has shown that proper PIA conditions can fully recover some of the unirradiated properties. However, the underlying process by which mitigation occurs is not well understood. Here, a commercial purity 304 stainless steel was irradiated to 10 dpa using 2 MeV protons at 360°C. Irradiated samples were annealed at different conditions: 500°C to 1, 15h, 100h, 550°C to 1h and 600°C for 1 and 10h. The irradiated and annealed microstructures were examined using transmission electron microscopy (TEM) and atom probe tomography (APT), revealing that radiation-induced Cu-rich and Ni/Si-rich clusters were present in CP304 after proton irradiation. The number density of Cu clusters decreased after PIA and completely dissolved after annealing at 600°C for 10h. The Ni-to-Si ratio in the Ni/Si clusters changed after PIA due to the difference in diffusivity of Si and Ni. The Ni/Si clusters dissolved after PIA at 600°C:10h. The details of microstructure quantification using APT will also be discussed.

Tungsten and tungsten alloys are among the most promising candidates for plasma-facing nuclear fusion reactor components due to their superior radiation resistance and high thermal conductivity. Fundamental understanding of the effects of radiation damage on the mechanical properties of such materials is essential for the design of next-generation fusion reactors. In this study we investigate the effect of helium-ion implantation on the elastic anisotropy of single-crystal (110) tungsten by means of impulsive stimulated thermal scattering (ISTS) spectroscopy. In ISTS, two pump laser pulses are crossed at the surface of the sample, generating a spatially periodic intensity profile. Impulsive thermal expansion due to this intensity grating generates counter-propagating surface acoustic waves (SAWs). These can be detected by subsequent diffraction of a probe beam from the time-dependent surface displacement profile. ISTS measurements were performed as a function of in-plane propagation direction on implanted and unimplanted (110) single crystal tungsten samples. Green’s tensor optimization procedures were used to determine the elastic constants, and from these the Zener anisotropy ratio, A, was calculated. Unimplanted tungsten is almost perfectly isotropic (A = 1.01). The helium-ion implantation causes a very slight increase in anisotropy (A = 1.02). Our results show that ion implantation can cause surprising changes in elastic properties. Indeed our observations are in remarkably good agreement with density functional theory calculations that previously predicted this effect. The ability of the ISTS technique to accurately measure such small anisotropy changes is a testament to its potential for high-precision mechanical characterization of a wide range of materials.

In this research we have studied ½<111>{110} edge dislocation interactions with rigid inclusions in bcc Fe. The inclusions were created from the same bcc materials as matrix Fe but having very large elastic moduli to simulate qualitatively the possible response from rigid oxide inclusions. A wide range of parameters such ad inclusion size and interaction geometry as well as ambient temperature and strain rate were studied with the main objectives focused on the direct comparison with available and future experiments. Stress-strain behavior, critical resolved shear stress and the interaction mechanisms and their temperature behavior were analyzed using a recently developed dislocation characterization and visualization technique. The high quality visualization of dislocation-obstacle interaction provides a direct comparison with in situ deformation TEM experiments and especially with the recently developed 3D TEM tomography.

This work was supported by the US Department of Energy Office of Fusion Energy Sciences.

Current first-generation candidate alloys that are envisioned to replace austenitic steels are ferritic and ferritic-martensitic steels that are known as a class to be more swelling-resistant than austenitics. Second-generation candidate alloys are oxide-dispersion-strengthened (ODS) variants of these ferritic and ferritic-martensitic steels. In these alloys nano-sized oxide particles or "dispersoids" stabilize grain boundary walls and thereby improve the resistance to both thermal creep and irradiation creep. When severe deformation techniques are used to generate very small size grains, oxide particles serve to maintain the size distribution and resist recrystallization, thereby preserving grain boundaries as sinks for radiation-produced point defects. Studies also show that the interfaces of oxide particles within the nano-grains also act as defect sinks to reduce radiation damage.
The effectiveness of dispersoids as grain stabilizers and defect sinks is dependent on size, density and uniformity of the particle distribution. In our studies using self-ion irradiation, oxide particles are observed to quickly evolve to an equilibrium size which depends primarily on irradiation temperatures and perhaps dpa rate. It is seen that the higher the irradiation temperature, the larger the equilibrium size. But the details of this temperature dependence behavior is complicated, involving the effects of interface coherence levels, defect creation rates, and damage cascade morphology.
In the present study, alloy engineering was used to prepare a range of ODS alloys having different sizes, densities and interface coherence levels, primarily through composition adjustments of Zr or Ti and Al. The ODS alloys are being irradiated by 3.5 MeV Fe self-ions at temperatures ranging from 100 to 600°C to systematically study both oxide stability and void swelling resistance. A physics-based model is proposed to explain the observed variations in radiation behavior with the ultimate goal of predicting ODS behavior under conditions outside the boundaries of the test matrix.

A mean-field kinetic model, implemented in the Radiation-Induced Microstructure Evolution (RIME) code, is used for understanding gas bubble evolution in the grain interior and on grain boundaries in high Ni alloys. The model for grain boundaries is entirely novel. The code accounts for the production of Frenkel pairs, with interstitial atoms ejected from and vacancies incorporated into the bubble, to reduce pressure of the bubble with high helium-to-vacancy ratio. The code tracks thermal neutron fluence dependence of He and atomic displacement production due to the two-step nuclear reaction with Ni. The calculations show that the bubble nucleation and growth on grain boundaries is different from those inside the grain. The results are compared with available data on Inconel X750 irradiated up to 100 dpa at 200 and 315°C. This reveals the necessity of inclusion of the effect of residual gas atoms on cavity nucleation to reduce the strong temperature dependence of bubble density that is otherwise predicted.

Intermetallic precipitates form in the oxide layer of zirconium alloys due to the low solubility of alloying elements in zirconium oxide. As they get oxidized, they form interfaces with ZrO₂ while being embedded into the passive film. It is known from literature that these secondary-phase particles do impact the corrosion and hydrogen pick-up resistance of zirconium alloys. However, it is not known how they locally and macroscopically affect the ionic and electronic transport properties of the passive film. Predicting the effect of such interfaces on oxidation and hydrogen pickup process is an important part of understanding the role of alloying elements in zirconium alloy degradation. We present a first-principle based method to predict the spatially-resolved defect concentrations across oxide/oxide hetero-interfaces, and implement that framework to the ZrO₂/Cr₂O₃ interface, which forms upon the oxidation of Zr(Fe,Cr)₂ intermetallic precipitate. Three factors are taken into consideration in the modeling framework: band discontinuities, defect segregation due to the local chemical and mechanical environment variation close to the interface, and drift-diffusion effect of charged defects in the space-charge layer. The first two are obtained from density functional theory calculations and the third from continuum-level modeling. This multi-scale method provides a quantitative evaluation of the effect of alloying elements on corrosion resistance when they precipitate as secondary phase particles. Our results show that in the extended space charge layer formed around the Cr₂O₃ particles in ZrO₂, the electronic conductivity and oxygen vacancy concentration are decreased by several orders of magnitude. This indicates a slower oxidation kinetics. Increase in electron concentration also promotes proton reduction and hydrogen evolution at the surface, and thus suppress hydrogen pickup into the alloy. Both effects are beneficial towards protecting the integrity of zirconium alloy. Our computational framework provides a physically-based approach to predicting how complex hetero-interfaces in the oxidation layer affect the corrosion kinetics, and how the passive layers can be engineered.

Kinetic Monte Carlo (KMC) has been known as a powerful method of computing the effective diffusion coefficients of transporting species with complicated diffusion paths and special heterogeneities. This talk presents an accelerated (KMC) method parameterized by density functional theory (DFT) calculations to compute the diffusivity of hydrogen in hcp metals, which involves multiple paths and jump frequencies in between the tetrahedron (T) and octahedron (O) interstitial sites. The T-T first nearest neighbor jump has a much lower barrier than any other jumps and forms a low barrier basin, strongly limiting the efficiency of KMC simulations. To overcome this challenge, analytical solutions are derived for the average escaping time and probability for each escaping path. The results are used to replace the regular KMC events to improve the simulation efficiency. The accelerated KMC has been applied for hydrogen diffusion in α-zirconium with the attempt rate and activation barrier associated with each jump obtained using density functional theory calculations. The accelerated KMC is shown capable of efficiently calculating the hydrogen diffusivity along and directions, without changing the kinetics of long-range diffusion. The results on hydrogen diffusivity are in good agreement with previous experiments and theories, and they further elucidate the controversy on the anisotropy of hydrogen diffusion. The accelerated KMC is shown capable of efficiently calculating the hydrogen diffusivity along and directions, without changing the kinetics of long-range diffusion. The results are in good agreements with previous experimental measurements and help to elucidate the anisotropy of hydrogen diffusion in α-zirconium and Zircaloy. In general, the method can be applied to any interstitial diffusion in hcp metals.

The integrity of fuel rods containing uranium oxide fuel pellets in a nuclear reactor core is of the utmost importance in pressurized water reactors (PWRs). One of the main threats is corrosion, specifically CRUD-induced localized corrosion (CILC) accelerated by the buildup of porous corrosion deposits (CRUD), which impede heat transfer and exaggerate water chemistry effects. Accurately predicting corrosion, including local effects such as CILC, would mark a significant improvement in PWR plant operation and safety. Cladding corrosion models are ubiquitous in the nuclear industry, but historically have been empirical in nature. Traditional models frequently do not possess the dimensional detail to capture effects on a local scale, such as CILC. The authors are developing a more science-based, mechanistic, finite-element model of zirconium alloy corrosion by investigating the corrosion of various alloys under a range of applicable conditions. By drawing from data from atomistic simulations and higher length scales, the highly coupled, multiphysics corrosion phenomena present in the cladding system can be modeled and investigated in a detailed manner previously untreated with such precision and underlying physics. Using a phase field model implemented in a finite-element framework, temperature, oxide, and oxygen distributions are determined in the cladding on the micron length scale.

This corrosion model, HOGNOSE (High-speed Oxide Growth in a Numerically Optimized Simulation Environment), incorporates oxygen transport through the metal and oxide, metal/oxide phase change, variations in oxide space charge with electron drift, and cracking and percolated porosity effects. Alloys vary in the time-dependence of their corrosion kinetics, and in contrast with other corrosion models, HOGNOSE does not assume any form of oxide thickness growth over time, such as imposing a cubic rate law. Rather, HOGNOSE accounts for variations in corrosion behavior by considering the effects of alloying elements on material properties related to electron, oxygen ion, and oxygen vacancy transport. As a result, HOGNOSE is able to accurately capture the initial kinetics of Zircaloy-4 before the kinetic transition and also addresses a key factor in irradiation-enhanced corrosion. HOGNOSE utilizes data from atomistic simulations and molecular dynamics in its treatment of oxygen transport in order to predict corrosion on the cladding length scale.

Waterside corrosion and associated hydrogen generation remain a major limiting factor for the use of zirconium alloys as cladding materials in the event of severe fuel duty conditions. Although it has been observed that corrosion rates strongly depend on minor alloying additions, the reasons for these differences remain unclear. In this study, we focus on the structure and chemistry of two different alloys: pure Zr and ZrFeCr alloy. The two alloys were corroded in 360°C water for 1 to 456 days and exhibited dramatically different corrosion behaviors. To understand the these differences, we used a multi-scaled characterization approach combining scanning electron microscopy, focused ion beam milling, transmission electron microscopy, and atom probe tomography. Complex oxide phase structures, morphologies, and grain textures were observed in the oxide scales formed on both alloys, along with enhanced inhomogeneity of alloying elements relative to base alloys. While ZrFeCr formed a uniform and stable oxide scale, pure Zr exhibited unstable oxide growth, forming oxide dendrites that penetrate farther into metal grain boundaries. By comparing the microstructures and chemistry of the two oxidized alloys, we identified factors that could potentially impact the protective character of the oxide phases, which include the distributions of alloying elements in the oxide phases and near the metal/oxide interface, grain boundary chemistry in the oxide scale and substrate metal, oxide grain morphologies and oxide textures. The contribution of each factor will be assessed and reviewed.

Zirconium alloys are the current canonical cladding material for the nuclear fuel. In nuclear reactor conditions, a zirconium oxide (ZrO₂) layer grows natively on zirconium alloys as a protecting passive film against further oxidation. The intactness and protectiveness of this ZrO₂ is crucial for the safe operation of nuclear reactors.
Because of its importance, many efforts have been to understand its chemistry, structure and electronic structure while growing on Zr alloys In our recent work, we started from zirconium metal single crystal as a model system and observed through scanning tunneling microscopy (STM) experiments, that the initial oxide layer growing on zirconium metal exhibited a distinctive triangular Moiré pattern and that the resulting oxide surface layer was highly defective. We proposed that the atomic structure corresponding for the observation of such patterns was cubic ZrO₂ oriented in <111> direction. Further we explained that the triangular Moiré pattern could come from the different stacking sequences of the initial oxide layer with respect to the underlying metal substrate and that the highly defective surface could be due to the formation of oxygen vacancies. [M1] However, the confirmation of this proposal required atomistic surface modelling. In this work we examined the first oxide layer that grows on Zr metal using density functional theory. We show that similar STM images can be generated computationally using first principles calculations on an oxide-metal interface consisting of an array of three zirconium metal layers underneath one bi-layer of zirconium oxide with the orientation of (111). It is found that the Moiré patterns originate from the relaxation of the oxide layer as well as the underlying metal layer, which is induced by the mismatch in the two layers' lattice parameters. This result is in a great agreement with our experimental observations. For a better understanding of the origin of the highly defective surface, we conduct several other computational calculations in the same interface structure related to the formation of oxygen vacancies. We find that some of the features found in the highly defective surface oxide layer in the STM images can be reproduced computationally by incorporating oxygen vacancies into the surface of the oxide layer of the interface. These results pave a way for interpreting oxidation properties of zirconium and making further investigations in the formation of its surface oxide.